ML20011D517: Difference between revisions

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| number = ML20011D517
| number = ML20011D517
| issue date = 12/15/1989
| issue date = 12/15/1989
| title = Amend F to C-E Std SAR - Design Certification (CESSAR-DC).
| title = Amend F to C-E Std SAR - Design Certification (CESSAR-DC)
| author name =  
| author name =  
| author affiliation = ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
| author affiliation = ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
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Enclosure II 13 LD-89-145 COMBUSTION ENGINEERING STANDARD SAFETY ANALYSIS REPORT -
DESIGN CERTIFICATION (CESSAR-DC) AMENDMENT F t
i i
                                                                          .~
l l
i f
8912270353 891222
    ., PDR  ADOCK 05000470 '
    ' A                PDR
 
C E S S A R RB h r.o.                              (Shoot 2 or S)
L EFFECTIVE PAGE LISTING CEAPTER 4 Table of Contents Ragg                                        Amendment l--    i 11                                                B iii iv v
vi vii viii                                              B ix x                                                F xi xii xiii                                              B xiv                                                B xv.                                                B xvi                                                B xvii                                              B xviii                                              F xix                                                F xx                                                B xxi                                                B xxii Text 23g3                                        Junendment 4.1-1                                            B 4.1-2                                            F 4.2-1                                              F 4.2-2                                              B 4.2-3                                              F 4.2-4 4.2-5 4.2-6 4.2-7                                              B 4.2-8                                              B 4.2-9 4.2-10 4.2-11 4.2-12 4.2-13                                            B 4.2-14                                            F                ;
Amendment F December 15, 1989
 
CESSARBR h                                                (Sn=t 2 or S)                      !
l i
EFFECTIVE PAGE LISTING (Cont'd)                                              I
!                              Chapter 4                                                          j l
13E1 (Cont'd)                                                                            i
!  Ragt                                                  Amendment I  4.2-15                                                    F 4.2-16 4.2.17 4.2-18                                                                                          1 4.2-19 4.2-20 4.2-21                                                    B 4.2-22                                                    F                                  !
4.2-23                                                                                          l 4.2-24                                                                                          !
4.2-25 4.2                                                    B 4.2-27                                                    F l
4.2-28                                                    F 4.2-29                                                    F                                  )
4.2-30                                                    F                                  :
1 4.2-31 4.2-32 4.2-33 4.2-34                                                    F 4.2-35                                                    F 4.2-36                                                    F 4.2-37                                                    B 4.2-38                                                    B 4.2-39                                                    B 4.2-40                                                                                        '
4.2-41                                                    B 4.2-42                                                    F 4.2-43.                                                    F 4.2-44                                                    F 4.2-45                                                    F 4.2-46                                                    F 4.2-47                                                    F 4.2-48                                                    B                                  -
4.2-49 4.2-50                                                    F 4.2-51                                                    F 4.2-52                                                    B
  > 4.2-53                                                    F 4.2-54 4.2-55                                                    F 4.2-56                                                    B 4.2-57 4.2-58                                                    F AinendInent F December 15, 1989
* GESSAOIB h                                                                  (shoot 3 or 9)
EFFECTIVE PAGE LISTING (Cont'd)
Chapter 4
_          Text (Cont'd)
Page                                                                  Amendment 4.2-59 4.2-60 4.2-61 4.2-62                                                                      B 4.2-63                                                                      B 4.2-64                                                                      B 4.2-65 4.2-66 4.2-67 4.2-68                                                                      F 4.2-69                                                                      F 4.2-70                                                                      F 4.2-71 4.2-72                                                                      F 4.2-73                                                                      B 4.2-74 4.2-75 4.2-76                                                                      B 4.2-77                                                                      B i        4.2-78                                                                      F 4.2-79                                                                      F 4.2-80                                                                      F 4.2-81                                                                      F 4.2-82 4.2-83                                                                      B l
4.2-84                                                                      F 4.2-85 4.2-86                                                                      F
(.
!        4.2-87                                                                      F 4.2-88                                                                      B 4.3-1                                                                        F 4.3-2 4.3-3 4.3-4                                                                        B 4.3-5                                                                        F 4.3-6                                                                        F l        4.3-7                                                                        F 4.3-8                                                                        B 4.3-9                                                                        B 4.3-10 l-      _4.3-11                                                                        B l        4.3-12                                                                      B l                                                                                                                          i Amendment F                        l December 15, 1989
                                                                . . _ , . _ - _ .,__ _ __...____-____.___ __ _ ..___ _ _ J
 
  \!'
CESSARtih                                                                                                      (sh::t 4 or 9)
I i
EFFECTIVE PME LISTENG (Cont'd)
Chapter 4                                                                              ,
2331 (Cont'd)                                                                                                              :
i Rage                                                                                        Amendment                            ,
4.3-13                                                                                                                          ;
4.3-14 4.3-15                                                                                                        B                ,
4.3-16                                                                                                        B                6 4.3-17                                                                                                        B 4.3-18                                                                                                        B 4.3-19                                                                                                        F 4.3-20 4.3-21 4.3-22                                                                                                        F 4.3-23                                                                                                        B 4.3-24                                                                                                        B                ,
4.3-25                                                                                                        B 4.3-26                                                                                                                          -
4.3-27                                                                                                        B 4.3-28                                                                                                        B 4.3-29                                                                                                        B 4.3-30                                                                                                        B 4.3-31                                                                                                        B 4.3-32                                                                                                        B
          -4.3-33 4.3-34                                                                                                        B 4.3-35 4.3-36                                                                                                        B 4.3-37                                                                                                        B 4.3-38                                                                                                        B 4.3-39                                                                                                        B 4.3-40 4.3-41                                                                                                        B 4.3-42                                                                                                        F 4.3-43                                                                                                        B 4.4-1                                                                                                          F 4.4-2 4.4-3                                                                                                          F 4.4-4                                                                                                          B
          '4.4-5                                                                                                          F 4.4-6                                                                                                          B 4.4-7                                                                                                          B 4.4-8                                                                                                          B 4.4-9                                                                                                          B 4.4-10                                                                                                        F 4.4-11 4.4-12 Amendment F December 15, 1989
 
                                        ~
      'CESSARBB h .
.                                                    (Shoot 5 or S) i:
RFFECTIVE PAGE LISTING (Conted) chapter _e M (conted)
@      Easa                                      pendment 4.4-13                                        B 4.4-14                                        B 4.4-15 4.4-16                                        F 4.4-17                                        B 4.4-18                                        F 4.4-19                                        F 4.4-20                                        B 4.4-21                                        B 4.4-22 4.4-23                                        F 4.4-24                                        B 4.4-25 4.4-26 4.4-27 4.4-28 4.4-29                                          B 4.4-30                                          B 4.4-31 4.4-32                                        B 4.4-33                                        B 4.4-34                                        F 4.4-35 4.4-36 4.4-37                                        B 4.4-38 4.4-39 4.5-1                                          F 4.5-2                                          F 4.5-3                                          D
  <    4.5-4 4.5-5                                          F 4.5-6                                          F 4.5-7                                          D 4.5-8                                          F 4.5-9 4.5-10                                          D 4.5-11 4.5-12 4.6-1                                          B 4.6-2                                          F 4.6-3                                          F Amendment F December 15, 1989
 
                    \
f9 s
            -CESSARtu b                                                                                                                        (Sh'ot 5 or S)                                      .
1 i
i EFFECTIVE PAGE_11522M9 (Conted)                                                                                                                              )
ghp.oter a Tables                                                                                                                      haandment
                                                                                                                                                                                                    ]
4.2-1 (Sheet 1)                                                                                                                  B                                                    l 4.2-1 (Sheet 2)                                                                                                                  F 4.2-1 (Sheet 3)                                                                                                                  F 4.2-1.(Sheet 4)                                                                                                                  F                                                  j 4.2-1 (Sheet 5)                                                                                                                                                                      :
4.2-1 (Sheet 6)                                                                                                                  B                                                  ;
4.2-2 4.2-3 (Sheet. 1)                                                                                                                  F                                                  :
4.2-3 (Sheet 2)                                                                                                                  F 4.3-1 (Sheets 1 and 2)                                                                                                            B                                                  ,.
4.3-2                                                                                                                            B 4.3-3                                                                                                                            B 4.3-4                                                                                                                            B 4.3-5                                                                                                                            B                                                  !
4.3-6                                                                                                                            B                                                  i 4.3-7                                                                                                                            B
            ' 4.3-8                                                                                                                            B 4.3-9 4.3-10                                                                                                                          B                                                  7 4.3-11                                                                                                                                                                              ;
4.3-12                                                                                                                                                                              ;
4.3-13                                                                                                                                                                              l 4.3-14                                                                                                                          B 4.3-15                                                                                                                          B 4.3-16 4.3-17 4.3-18                                                                                                                                                                              -
4.3-19 4.4-1 (Sheet 1)                                                                                                                  F 4.4-1 (Sheet 2)                                                                                                                  F                                                  :
4.4-2                                                                                                                            F 4.4-3 4.4-4                                                                                                                            F 4.4-5 4.4-6                                                                                                                            F 4.4-7 4
4.4-8                                                                                                                            F                                                  >
:.            4.4-9 (Sheet 1)                                                                                                                  F 4.4-9 (Sheet 2)                                                                                                                  B 4.6-1                                                                                                                            B                                                  '
i Amendment F December 15, 1989
 
      . CESSARRB h                                                                                                            (Sh00t 7 of 9) l l
i EFFECTIVE PAGE LISTING (Conted)                                                                                      l 1
chapter 4                                                                  ;
                                                                                                                                                                                )
Figures                                                                                                        Amendment                                              >
4.1-1                                                                                                                  F                                                  ,
4.1-2                                                                                                                  F                                                  l 4.2-1                                                                                                                                                                    .
4.2-2                                                                                                                                                                    i 4.2-3                                                                                                                  F                                                !
4.2-4                                                                                                                  B                                                ;
4.2-5                                                                                                                  F 4.2-6                                                                                                                  B 4.2-7                                                                                                                  B                                                i 4.2-8                                                                                                                  F i      4.2-9                                                                                                                  F 4.2-10                                                                                                                F                                                ;
B 4.2-11                                                                                                                                                                  '
4.3-1                                                                                                                  B                                                ,
4.3-2                                                                                                                  B t
4.3-3                                                                                                                  B                                                ,
I      4.3-4                                                                                                                  B 4.3-8                                                                                                                  B                                                i 4.3-Sa                                                                                                                B                                                i 4.3-6                                                                                                                  B 4.3-6a                                                                                                                B 4.3-7                                                                                                                  B                                                ,
4.3-7a                                                                                                                B                                                .
4.3-8                                                                                                                  B 4.3-8                                                                                                                  B 4.3-9a                                                                                                                B 4.3                                                                                                                B.
j'      4.3-10a                                                                                                                B 4.3-11                                                                                                                B 4.3-11a                                                                                                                B 4.3-12                                                                                                                B 4.3-13'                                                                                                                B                                                ,
4.3-14                                                                                                                B i-B 4.3-15 4.3-16                                                                                                                B 4.3-17                                                                                                                B                                                ,
4.3-18                                                                                                                B 4.3-19                                                                                                                B 4.3-20                                                                                                                B                                                >
4.3-21                                                                                                                B 4.3-22                                                                                                                B 4.3-23                                                                                                                B 4.3-24                                                                                                                F 4.3-25                                                                                                                F I
Amendment F l                                                                                                                              December 15, 1989
  . ..  . _ _ . _ _ . . . . _ . _ . . _ _ _ _ _ _ . . . . . _ . _ . . _ _ _ _ . . ~ . _ _ _ _ ~ . . . . ~ . _              __        . _ _ . _ _ _ _ . _ . _ _ . . _ . -
 
      'CESSARBR h                                          (Sh::t 8 or $)                            ,
i s                                                                                                    i i
I l
EFFECTIVE PAGE LISTING (Cont'd)
Chapter 4 Figures (Cont'd)                            Amendment 4.3-26                                            F                                          1 4.3-27                                            F                                          i 4.3-28                                            F                                          l 4.3-29                                            F.
4.3-30                                            F                                        .1 4.3-31                                            F                                          >
4.3-32                                            F                                          l 4.3=33                                            F 4.3-34                                            F                                          '
4.3-35                                            F 4.3-36                                            E                                        I 4.3-36a                                            E                                          ,
4.3-37                                            E                                          :
4.3-37a                                            E 4.3-38                                            B 4.3-39                                            B 4.3-40                                            B 4.3-41                                            B i      4.3-42                                            B t      4.3-43                                            B 4.3-44                                            B                                          ,
4.3-45                                            B 4.3                                            B                                        '
;      4.3-47                                            B 4.3-48                                            B 4.3-49                                            E                                        ,
4.3-50                                            B i      4.3-51                                            B 4.3-52                                            B 4.3-53                                            B 4.3-54                                            B
;. 4.3-55                                            B 4.3-56                                            B                                        ,
4.3-57                                            B 4.3-58                                            B 4.3-59                                            B 4.3-60                                            B
:      4.3-61                                            B 4.3-62                                            B
  . 4.3-63                                            B l    4.4-1                                              B 4  4.4-2                                              B
                      ~
i Amendment F
.                                                        December 15, 1989
: s.                                                                                                  .
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7
          ' CESSARIBrb                                                                                                            (Sh=st $ cr S) l l
EFFECTIVE PAGE LISTING (Cont'd) chapter 4 Figures (Cont'd)                                                                                                  naendment i
4.4-3                                                                                                                  B                              l 4.4-4                                                                                                                  B                              <
4.4-5                                                                                                                  B                              l' 4.4-6                                                                                                                  F 4.4-7                                                                                                                  B 4.4-8                                                                                                                  B i
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l Amendment F                  ,
December 15, 1989 I
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                                                                                                                                                  , . - - a - ,
 
P CESSARtR A ThBLE OF COMTENTS CEAPTER 4 section    subject                                  Pane No.
4.0        REACTOR                                  4.1-1 4.1       
 
==SUMMARY==
DESCRIPTION                      4.1-1 4.2        FUEL SYSTEM DESIGN                        4.2-1 4.2.1      DESIGN BASES                              4.2-1 4.2.1.1          Fuel Assembly                        4.2-1 4.2.1.1.1        Fuel Assembly Structural            4.2-3 Integrity Criteria 4.2.1.1.2        Material Selection                  4.2-6 4.2.1.1.3        Control Element Assembly            4.2-6 Guide Tubes 4.2.1.1.4        Zircaloy-4 Bar Stock                4.2-7 4.2.1.1.5        Zircaloy-4 Strip Stook              4.2-7 4.2.1.1.6        Stainless Steel Castings            4.2-8 4.2.1.1.7        Stainless Steel Tubing              4.2-D 4.2.1.1.8        Inconel X-750 Compression          4.2-8 Springs 4.2.1.1.9        Inconel 625 Bottom Spacer Orid      4.2-8 Strip Material 4.2.1.2          Fuel Rod                            4.2-9 4.2.1.2.1        Fuel Cladding Design Limits        4.2-9 4.2.1.2.2        Fuel Rod Cladding Properties        4.2-12      i 4.2.1.2.2.1            Mechanical Properties        4.2-12 4.2.1.2.2.2            Dimensional Requirements      4.2-13 4.2.1.2.2.3            Metallurgical Properties      4.2-13 4.2.1.2.2.4            Chemical Properties          4.2-13 4.2.1.2.3        Fuel Rod Component Properties      4.2-13 4.2.1.2.3.1            Zircaloy-4 Bar Stock          4.2-13 4.2.1.2.3.2            Stainless Steel Compression  4.2-14 Springs 4.2.1.2.4        UO 2 Fuel Pellet Properties        4.2-14 i
 
CESSAR88W, TABLE OP CONTENTS CEAPTER 4 Section              Subiect                            Pace No.
4.2.1.2.4.1              Chemical Composition          4.2-14 4.2.1.2.4.2              Microstructure                4.2-15 4.2.1.2.4.3              Density                        4.2-15 4.2.1.2.4.4              Thermal Properties            4.2-16 4.2.1.2.4.5              Mechanical Properties          4.2-17 l
4.2.1.2.5            Fuel Rod Pressurization            4.2-17 4.2.1.2.5.1              Capacity for Fission Gas      4.2-18 Inventory 4.2.1.2.5.2              Fuel Rod Plenum Design        4.2-19 4.2.1.2.5.3              Outline of Procedure Used      4.2-19 to Size the Fuel Rod Plenum 4.2.1.2.6            Fuel Rod Performance                4.2-21 4.2.1.3              Burnable Poison Rod                4.2-21 4.2.1.3.1            Burnable Poison Rod Cladding        4.2-21      ,
Design Linits 4.2.1.3.2            Burnable Poison Rod Cladding        4.2-22 Properties 4.2.1.3.3          Al O ~B C Burnable Poison          4.2-22 Pe$l$t hroperties                              -
4.2.1.3.3.1              Thermal-Physical Properties    4.2-22 4.2.1.3.3.2              Irradiation Properties        4.2-24 4.2.1.3.3.3              Chemical Properties            4.2-25 0
4.2.1.3.4          Gd  -UO  Burnable Poison Pellet  4.2-26 l                        Pr$drti$s 4.2.1.4            Control Element Assemblies (CEAs)  4.2-26 4.2.1.4.1          Thermal-Physical Properties of      4.2-27 Absorber Material 4.2.1.4.2          Compatibility of Absorber and      4.2-30 Cladding Materials 4.2.1.4.3          Cladding Stress-Strain Limits      4.2-30 4.2.1.4.4          Irradiation Behavior of Absorber    4.2-32 Materials Amendment B 11                March 31, 1988
 
l C E S S A R RB M.eu l
ThBLE OF CONTENTS CRAPTER 4                                1 l
Section        Bubient                                  Pace No.    :
4.2.1.5              Surveillance Proaram                4.2.35 l
4.2.1.5.1            Requirements for Surveillance        4.2-35 l
and Testing of Irradiated Puel Rods 4.
 
==2.2          DESCRIPTION==
AND DESIGN DRAWINGS          4.2-36 1
4.2.2.1              Fuel Assembly                        4.2-37 4.2.2.2              Fuel Rod                            4.2-40 4.2.2.3              Burnable Poison Rod                  4.2-41      {
4.2.2.4              Control Element Assembiv            4.2-42 Descriorion and Desian 4.2.3          DESIGN EVALUATION                        4.2-44 4.2.3.1              Fuel Assembiv                        4.2-44 l
l    4.2.3.1.1            Vibration Analyses                  4.2-44 4.2.3.1.2            CEA Guide Tube                      4.2-45 4.2.3.1.2.1                Operating Basis Earthquake    4.2-46 l                              (OBE)
;    4.2.3.1.2.2                Safe Shutdown Earthquake      4.2-46 (SSE)
E 4.2.3.1.2.3                Loss-of-Coolant Accident      4.2-46
,                              (LOCA) l    4.2.3.1.2.4                Combined SSE and LOCA          4.2-46 l
4.2.3.1.3            Spacer Grid Evaluation              4.2-47 4.2.3.1.4            Dimensional Stability of Zircaloy    4.2-48 4.2.3.1.5            Fuel Handling and Shipping          4.2-49 Design Loads 4.2.3.1.6            Puel Assembly Analysis Results      4.2-50 4.2.3.1.7            Fuel Assembly Liftoff Analysis      4.2-50 1
l iii l
o l ..    --      --      --        - - . - -
 
CESSAR tm            no.
I i
I T1 ELE OP CONTENTS CEAPTER 4 Section        Subient                                              Pace No.
i 4.2.3.2              Puel Rod Deslan Evaluation                      4.2-50                l 4.2.3.2.1            Results of Vibration Analyses                    4.2-50 4.2.3.2.2            Fuel Rod Internal Pressure and                  4.2-50 Stress Analysis 4.2.3.2.3            Potential for Chemical Reaction                  4.2-51 4.2.3.2.4            Fretting Corrosion                              4.2-53 4.2.3.2.5            Fuel Rod Bowing                                  4.2-53 4.2.3.2.6            Irradiation Stability of Fuel                    4.2-54                )
Rod Cladding                                                            i 4.2.3.2.7            Cladding Collapse Analysis                      4.2-55                l 4.2.3.2.8            Fuel Dimensional Stability                      4.2-55                )
4.2.3.2.9            Potential for Waterlogging                      4.2-56                :
Rupture and Chemical Interaction                                        j 4.2.3.2.10          Fuel Burnup Experience                          4.2-57                l 4.2.3.2.11          Temperature Transient Effects                    4.2-64                l Analysis                                                                ]
4.2.3.2.11.1              Waterlogged Fuel                          4.2-64 4.2.3.2.11.2              Intact Fuel                                4.2-65 4.2.3.2.12-          Energy Release During Fuel                      4.2-65 Element Burnout 4.2.3.2.13          Energy Release on Rupture of                    4.2-65 Waterlogged Fuel Elements Fuel Elements 4.2.3.2.14          Fuel Rod Behavior Effects from                  4.2-66 Coolant Flow Blockage 4.2.3.2.15          Fuel Temperatures                                4.2-67 4.2.3.3              Burnable Poison Rod                              4.2-67 4.2.3.3.1            Burnable Poison Rod Internal                    4.2-67 Pressure and Cladding Stress 4.2.3.3.2            Poter;tial for Chemical Reaction                4.2-68 4.2.3.4              Control Element Assembly                        4.2-68 1
l 4.2.4          TESTING AND INSPECTION PLAN                          4.2-71 4.2.4.1              Fuel Assembly                                    4.2-71 iv
 
CESSARtRLn..
TkBLE OF CONTENTS                                                ;
CEAPTER 4                                                    l section            subient                                                            Page No.      ,
4.2.4.1.1                      Wald Quality Assurance Measures                        4.2-72 4.2.4.1.2                      Other Quality Assurance Measures                      4.2-73        -
4.2.4.2                        Fuel Rod                                              4.2-73        l 4.2.4.2.1                      Fuel Pellets                                          4.2-73 4.2.4.2.2                      Cladding                                              4.2-74 4.2.4.2.3                      Fuel Rod Assembly                                      4.2-75 4.2.4.2.3.1                          Stack Length Gage                                4.2-75 4.2.4.2.3.2                          Fluoroscopy                                      4.2-75        t 4.2.4.3                        Burnable Poison Rod                                    4.2-76        !
4.2.4.3.1                      Burnable Poison Pellets                                4.2-76 4.2.4.3.2                      Cladding                                              4.2-77 4.2.4.4                        Control Element Assenbligg                            4.2-78 4.2.5            REACTOR INTERFACE REQUIREMENTS                                      4.2-79        ,
4.3                NUCLEAR DESIGN                                                    4.3-1 4.3.1              DESIGN BASES                                                      4.3-1 4.3.1.1                        Excess Reactivity and Fuel Burnuo                      4.3-1 4.3.1.2                        Core Desian Lifetime and Fuel                          4.3-1 Reo3acement Procram 1
4.3.1.3                        Necative Reactivity Feedback                          4.3-1 4.3.1.4                        Reactivity Coefficients                                4.3-1 4.3.1.5                        Burnable Poison Reauirements                          4.3-2 4.3.1.6                        Stability criteria                                    4.3-2 4.3.1.7                        Maximum Controlled Reactivity                          4.3-2 Insertion Rate v
 
CESSAO tinh                                                                                              :
ThBLE OF COMTENTS                                                              ,
CEAPTER 4 Section      subient                                                                Pace No.
4.3.1.8            Power Distribution Control                                      4.3-2 4.3.1.9            Excess CEA Worth with Stuck Rod                                  4.3-3 Criteria l
4.3.1.10            Chemical Shim Control                                            4.3-3 4.3.1.11            HRximum CEA SDeeds                                              4.3-3              '
4.
 
==3.2        DESCRIPTION==
4.3-3              ,
4.3.2.              Nuclear Desian Descriotion                                      4.3-3 4.3.2.2            Power Distribution                                              4.3-4 4.3.2.2.1          General                                                          4.3-4 4.3.2.2.2          Nuclanr Design Limits on the                                    4.3-5 Power Distribution                                                                  i 4.3.2.2.3          Expected Power Distributions                                    4.3-6 4.3.2.2.4          Allowances and Uncertainties on                                  4.3-8 Power Distributions 4.3.2.2.5          Comparisons Between Limiting                                    4.3-8 and Expected Power Distributions l  4.3.2.3            Reactivity Coefficients                                          4.3-9 i
4.3.2.3.1          Fuel Temperature Coefficient                                    4.3-10 4.3.2.3.2          Moderator Temperature Coefficient                                4.3-10 4.3.2.3.3          Moderator Density Coefficient                                    4.3-11 l  4.3.2.3.4          Moderator Nuclear Temperature                                    4.3-11 Coefficient 4.3.2.3.5          Moderator Pressure Coefficient                                  4.3-12 4.3.2.3.6          Moderator Void Coefficient                                      4.3-12 4.3.2.3.7          Power Coefficient                                                4.3-12 4.3.2.4            Control Reauirements                                            4.3-14 4.3.2.4.1          Reactivity Control at BOC and EOC                                4.3-15 4.3.2.4.2          Power Level and Power                                            4.3-15              ,
Distribution Control 4.3.2.4.3          Shutdown Reactivity Control                                      4.3-16 vi
 
                                                                                      )
l CESSARIH Ln..                                                                    I l
I l
l ranLa or corrames                                      J CEAPTER 4                                        I section        Subital                                            Pace No.
i 4.3.2.4.3.1                Fuel Temperature Variation            4.3-16          i 4.3.2.4.3.2              Moderator Temperature                  4.3-16        l Variation                                              J 4.3.2.4.3.3              Moderator Voids                        4.3-17 4.3.2.4.3.4                Control Element Assembly Bite 4.3-17                  i 4.3.2.4.3.5              Accident Analysis Allowance            4.3-17        .
4.3.2.4.3.6              Available Reactivity Worth            4.3-17 4.3.2.5              Control Element Assembiv                    4.3-18 Patterns and Reactivity Worths 4.3.2.6              Criticality of Reactor Durina                4.3-19 Refuchfug 4.3.2.7              Stability                                    4.3-19 4.3.2.7.1            General                                      4.3-19        '
4.3.2.7.2            Methods of Analysis                          4.3-20 4.3.2.7.3            Expected Stability Indices                  4.3-21 4.3.2.7.3.1                Radial Stability                      4.3-21 4.3.2.7.3.2                Azimuthal Stability                    4.3-21 4.3.2.7.3.3                Axial stability                        4.3-22 4.3.2.7.4            Control of Axial Instabilities              4.3-22 4.3.2.7.5            Summary of Special Features                  4.3-22 Required by Xenon Instability 4.3.2.7.5.1                Features Provided for                  4.3-23 Azimuthal Xenon Effects 4.3.2.7.5.2                Features Provided For Axial            4.3-23 Xenon Effects and Power Distribution Effect and Control 4.3.2.8              Vessel Irradiation                          4.3-23 4.3.3          ANALYTICAL METHODS                                4.3-23 4.3.3.1              Reactivity and Power Distribution            4.3-23 4.3.3.1.1            Methods of Analysis                          4.3-23 vii
 
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CESSAR tiMien                                                                            !
TABLE.QF_COMTENTE                                          j CEAPTER 4                                            ;
Section        Sub4ect                                                Page Me.
4.3.3.1.1.1                    Cross Section Generation                4.3-24 4.3.3.1.1.2                    Coarse-Meeh Methods                      4.3-27    B      I 4.3.3.1.1.3                    Fine-Mesh Methods                        4.3-30          :
4.3.3.1.1.4                    Other Analysis Methods                  4.3-30          i 4.3.3.1.2                Comparisons with Experiments                  4.3-34 4.3.3.1.2.1                    Critical Experiments                    4.3-34 4.3.3.1.2.2                    Power Reactors                          4.3-36 4.3.3.1.9. 2.1                      Startup Data                      4.3-36          !
4.3.3.1 . 2.2                        Depletion Data                    4.3-38          *
      . 4. 3. 3. :  '3
                          .                      Assembly Power                              B Distributions                      4.3-39          ,
4.3.3.2                  Soatial Stability                              4.3-39 4.3.3.2.1                Methods of Analysis                            4.3-39
        . 4.3.3.2.2                Radial Xenon Oscillations                      4.3-40          '
4.3.3.2.3                Azimuthal Xenon Oscillations                  4.3-40          '
4.3.3.2.4                Axial Xenon Oscillations                      4.3-41 4.3.3.3                  Reactor Vessel Fluence Calculation 4.3-41 Model 4.4            THERMAL AND HYDRAULIC DESIGN                            4.4-1 4.4.1          DESIGN BASES                                            4.4-1 4.4.1.1                  Minimum DeDarture from Nucleate                4.4-1 Boilina Ratio (DNBR) 4.4.1.2                  Hydraulic Stability                            4.4-1 4.4.1.3                  Puel Desian Bases                              4.4-1 4.4,1.4                  Coolant Flow. Velocity, and Void              4.4-2 Fraction 4.
 
==4.2          DESCRIPTION==
OF THERMAL AND HYDRAULIC                    4.4-3 DESIGN OF THE REACTOR CORE 4.4.2.1                  Summary Comoarison                            4.4-3 Amendment B viii                          March 31, 1988
 
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i i
        ,e TABLE OF CONTENTS
      .,                                                                                                              I CEAPTER 4 Section      subient                                                Pace No.
4.4.2.2            Critical Heat Flux Ratios                        4.4-3 r
14.4.2.2.1          Departure from Nucleate Boiling                  4.4-3
    ,                                Ratio 4.4.2.2.2          Application of Power Distribution                4.4-4 and Engineering Factors 4'.4.2.2.2.1'            Power Distribution Factors                4.4-5
;'              o4.4.2.2.2.2              Engineering Factors                        4.4-7 L,-
4.4.2.2.3          Fuel Densification Effect on DNBR                4.4-8 14.4.2.3            Linear Heat' Generation Rate                    4.4-8 4.4.2.4            _ Void-Fraction Distribution                      4.4-8                          '
4.4.2.5'            Core Coolant Flow Distribution                  4.4-8                          -
                ~4.4.2.6            Core Pressure Droos and Hydraulic                4.4-10 L                                    Loads 4.4.2.6.1          Reactor Vessel Flow Distribution                4.4-10 4.4.2.6.2          Reactor Vessel and Core Pressure                4.4-10 Drops      .
4.4.2.6.3          Hydraulic Loads on Internal                      4.4-10 Components 4.4.2.7            Correlations and Physical Data                  4.4-11 4.4.2.7.1          Heat Transfer Coefficients                      4.4-11 4.4.2.7.2          Core Irrecoverable Pressure Drop                4.4-13 Coefficients 4.4.2.7.3          Void Fraction Correlations                        4.4-13 4.4.2.8            Thermal Effects of Ooerational                  4.4-13 Transients 4.4.2.9-          Uncertainties in Estime, tag                      4.4-14 4.4.2.9.1          Pressure Drop Uncertainties                      4.4-14 4.'4.2.9.2          Hydraulic Load Uncertainties                    4.4-14 ix
 
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i TABLE OF. CONTENT 4 CEAPTER 4 Section              Subiect                                      Pace No.
4.4.2.9.3                      Fuel and Clad Temperature          4.4-14      -
Uncertainty 4.4.2.9.4                      DNBR Calculation Uncertainties      4.4-14 B
4.4.2.9.5                      Statistical Combination of          4.4-16 Uncertainties (SCU) 4.4'2.10                      Flux Tilt Considerations            4.4-17 4.
 
==4.3                DESCRIPTION==
OF THERMAL AND HYDRAULIC        4.4-17
                          ' DESIGN OF THE REACTOR COOLANT SYSTEM (RCS) 4'.4.3.1                      Plant Conficuration Data            4.4-18 F
4.4.3.2'                      Operatina Restrictions on PumQn    4.4        4.4.3.3                        Temeerature-Power Ooeratina Man    4.4-19 4.4.3.4                        Load Followin0 Characteristics      4.4-19
  ~
4.4.3.5                        Thermal and' Hydraulic              4.4-19 Characteristics Table 4.4.4                EVALUATION                                  4.4-20 4.4.4.1                      -Critical Heat Flux                  4.4-20 4.4.4.2                        Reactor Hydraulics                  4.4-23 4.4.4.2.1                      Reactor Flow Model Tests            4.4-23 4.4.4.2.2                      Component Testing                  4.4-25 4.4.4.2.3                      Core Pressure Drop Correlations    4.4-27 4.4.4.3                        Influence of Power Distributions    4.4-29 4.4.4.4                        Core Thermal Resnonse              4.4-30 4.4.4.5                        Analvtical Methods                  4.4-30 Amendment F x                December 15, 1989
 
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TABLE OF COMTENTS                        l i
CHAPTER 4 Section                                  Subiect                                Pace No.
4.4.4.5.1-                                    Reactor Coolant System Flow      4.4-30 Determination 4.4.4.5.2                                    Thermal Margin Analysis          4.4-32 4.4.4.5.3                                    Hydraulic Instability Analysis    4.4-35 4.4.5                                    TESTING AND VERIFICATION              4.4-36 4.4.6                                    INSTRUMENTATION REQUIREMENTS          4.4-36 4.5                                      REACTOR MATERIALS                      4.5-1 4.5.1                                    CONTROL ELEMENT DRIVE STRUCTURAL      4.5-1 MATERIALS 4.5.1.1                                      Material Soecifications          4.5-1 L                      4.5.1.2                                      Control of the Use of 90 ksi      4.5-3        ;
lield Strenoth Material
: t. 5.1.3                                      control of the Uselof Sensitized  4.5-4 l                                                                    Austenitic Stainless Steel 4.5.1.3.1                                    Solution Heat Treatment          4.5-4 Requirements 4.5.1.3.2                                    Material Inspection Program      4.5-4 4.5.1.3.3                                    Avoidance of Sensitization        4.5-4 4.5.1.4                                      Control of Delta Ferrite in      4.5-5 l                                                                    Austenitic Stainless Steel Welds
                        '4.5.1.5                                      Cleanina and Contamination        4.5-5 l                                                                    Protection Procedures 4.5.2                                    REACTOR INTERNAIS MATERIALS            4.5-6 4.5.2.1                                      Material Soecifications          4.5-6 4.5.2.2                                      Weldina Accentance Standards      4.5-8 4.5.2.3                                      Fabrication and Precegsina of    4.5-8 i                                                                    Austenitic Stainless Steel l
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l-TABLE OF CONTENTS                                      i l                              CHAPTER 4                                        =i Section    Bubiect                                      Pace No.          +
4.5.2.3.1        Control of the Use of Sensitized        4.5-8 Austenitic Stainless Steel i
4.5.2.3.1.1            Solution Heat Treatment            4.5-9            ,
Requirements 4.5.2.3.1.2            Material Inspection Program        4.5-9 4.5.2.3.1.3            Unstabilized Austenitic            4.5-9 Stainless-Steels                                    !
;-    4.5.2.3.1.4            Avoidance of Sensitization        4.5-10 4.5.2.3.1.5            Retesting Unstabilized            4.5-11
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l                            Austenitic Stainless Steels l_
Exposed to Sensitizing r                        Temperature 4.5.2.3.2        Non-Metallic Thermal Insulation          4.5-11 L    4.5.2.3.3        Control of Delta Ferrite in Welds        4.5-11            4 i
4.5.2.3.4        Control of Electroslag Weld              4.5-11 p                      Properties l-4.5.2.3.5        Welder Qualification for Areas of        4.5-11            <
Limited Accessibility b    -4.5.2.4          Contamination Protection and            4.5-12            ,
L                      Cleanina of Austenitic Stainless Steel l    4.6        FUNCTIONAL DESIGN OF REACTIVITY              4.6-1 l                CONTROL SYSTEMS l
4.6.1      INFORMATION FOR THE CONTROL ROD              4.6-1 DRIVE SYSTEM (CRDS) 4.6.2      EVALUATION OF THE CRDS                        4.6-1 4.6.2.1          Sinale Failure                          4.6-1 i
4.6.2.2          Isolation of the CRDS from              4.6-1 Other Eculoment e
4.6.2.3          Protection from Common Mode              4.6-2 Failure I                                  xii l:
 
: C E S S A R BR anea m .
TABLE OF OONTENTS CHAPTER 4 Section        Subi9_g.h                              Pace No.
      .4.6.2.3.1            Pipe Breaks                      4.6-2 o
4.6.3          TESTING AND VERIFICATION OF THE CRDS    4.6-3 4.6.4          INFORMATION FOR COMBINED PERFORMANCE    4.6-3 OF THE REACTIVITY CONTROL SYSTEMS 4.6.5          EVALUATION OF COMBINED PERFORMANCE      4.6-3 APPENDIX 4A    SYSTEM 80 REACTOR FLOW MODEL TEST      4A-1 PROGRAM B
APPENDIX 4B    HOT LOOP FLOW TESTING OF SYSTEM 80      4B-1 FUEL AND CEA COMPONENTS Amendment B xiii    March 31, 1988
 
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LIST OF~ TABLES CHAPTER 4 table                subiect          s                                                      ,
4.2-L-                Mechanical Design Parameters 4.2-2                Tensile Test Results on Irradiated Baxton Core-III:
                ,                              Cladding 4.2-3                C-E Poolside-Fuel Inspection Program Summary 4.3-1                Nuclear Design Characteristics L                        4.3-2                Effective Multiplication Factors and Reactivity l~                                            Data 4            4.3-3'              _
Comparison of Core Reactivity Coefficients with                          ;
Those Used in Various Safety Analyses                                    '
4.3-4                Reactivity Coefficients 4.3-5                Worths of CEA Groups                                                    t k
N-423-6                CEA Reactivity Allowances
            ^
                      '4.3-7                  Comparison of Available CEA Worths and-Allowances L
                      -4.3-8                - Comparison of Rodded and Unrodded Peaking Factors                        i i                                            'for Various Rodded Configurations l
4.3-9                Calculated Variation of the Axial Stability Index During-the First Cycle                                            B 4.3-10                Control Element-Assembly Shadowing Factors 4.3-11                C-E Criticals 4.3-12                Fuel Specifications (KRITZ Experiments)                                  ;
4.3-13                Comparison of Reactivity Levels for Non-Uniform Core 4.3-14                BOC,  HZP,    Xe      Free,    Unrodded    Critical    Boron Concentration 4.3-15                ITC Summary for ROCS /DIT 4.3-16                Comparison of Control Rod Bank Worths Amendment B xiv                    March 31, 1988
 
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e LIST OF TABLES (Cont'6)                                                                                          {
CRAPTER 4 Table  Subiect 4.3-17  Comparison'of Power coefficients 4.3 Summary of ROCS /DIT Calculative Uncertainties B
4.3-19  Axial Xenon Oscillations 4.4-1  Thermal and Hydraulic Parameters 4.4-2  Comparison of the Departure from Nucleate Boiling Ratios Computed with-Different Correlations 4.4-3  Best Estimate Reactor Coolant Flows in. Bypass Channels 4.4-4  Reactor Vessel Best Estimate Pressure Losses and-Coolant Temperatures 4.4-5  Design Steady State Hydraulic                                      Loads      on' Vessel Internals and= Fuel Assemblies 4.4-6  RCS-Valves and Pipe Fittings 4.4-7  RCS' Design Minimum Flows 4.4-8  Reactor Coolant System Geometry 4.4-9  Reactor Coolant          System- Component                                Thermal      and Hydraulic Data 24.6-1  Postulated Accidents Amendment B xv                                              March 31, 1988
 
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        'A LIST OF FIGURES
* CEAPTER 4                                                    l Figure          Subiact 4.1-1          ' Reactor Vertical Arrangement 4.1-2            Reactor Core Cross Section - 241 Fuel Assemblies                                B 4.2-1            Circumferential Strain vs Temperature 4.2-2            Design Curve for Cyclic Strain Usage of Zircaloy-4                                  <
at 700*P                                                                        g  ,
4.2-3            Full Strength Control Element Assembly (4-Element)-                                .
4.2-4            Full        Surength          Control                Element    Assembly.
(12-Element)'
          ' ~
4i2-5            Part Strength Control Element Assembly
                      '.2-6 4                Fuel Assembly 4.2-7            Fuel Spacer Grid Overview L                      4.2-8            Fuel Rod c
4.2-9 A123 0 ~"4C Burnable Poison ~ Rod                                                  g 4.2-10          UO2 Gd Op3 Burnable Poison Rod                                                      e 4.2-11          Control Element Assembly Locations g                      4.3-1            First Cycle Fuel Loading Pattern 4.3-2            First Cycle Assembly Fuel Loading Waterhole and Shim Placement 4.3-3            Planar Average Power Distribution, Unrodded, BOC, No Xenon Full Power 4.3-4            Planar Average Power Distribution, Unrodded,                                          l Full Power, Equilibrium Xenon, 50 mwd /T 4.3-5            Planar Average Power Distribution, Unrodded,                                          l Full Power, Equi? ibriur Xenon, 1000 mwd /T
  "                    4.3-Sa          Planar      Average        Power              Distribution,    Bank    P2  B      l 4
Inserted,            Full      Power,                Equilibrjum    Xenon,            i 1000 mwd /T                                                                          j Amendment B                I xvi                              March 31, 1988            l
 
LV' CESSAR HWim..                                                                                ,
L                                                                                                {
LIST OF FIGURES (Cont'd)
CHAPTER 4 Figure                    Subiect 4.3-6'                    Planar Average Power Distribution, Unrodded,                      '
Fuel Power, Equilibrium Xenon, 6000 mwd /T 4.3-6a                    Planar    Average    Power    Distribution,    Bank    P2 Power,    Equilibrium    Xenon,    B
                                ' Inserted,    Full 6000 mwd /T 4.3                    Planar Average Power Distribution, Unrodded, Full Power, Equilibrium Xenon, 9000 mwd /T 4.3-7a                    Planar    Average    Power    Distribution,    Bank    P2 B
Inserted,    Full    Power,    Equilibrium    Xenon, 9000 mwd /T L    4.3-3                      Planar Average Power Distribution, Unrodded, Full Power, Equilibrium Xenon, 16500 mwd /T
                                                                                                  .t 4.3-9                    . Planar Average. Power. Distribution, Lead Regulating Bank Fully Inserted, Full Power, Equilibrium Xenon, 2000 mwd /T
!    4.3-9a                    Planar ' Average Power Distribution, Bank 3 and P2 L                                Fully Inserted, Full Power Equilibrium Xenon, 2000 mwd /T p    4.3-10                    Planar Average Power Distribution, Bank 3 Fully l
Inserted,    Full    Power,    Equilibrium    Xenon, 9000 mwd /T 4.3-10a                    Planar Average Power Distribution, Bank 3 and P2 Fully- Inserted, Full Power Equilibrium Xenon,              B 9000 mwd /T                                                          l 4.3-11                    Planar Average Power Distribution, Bank 3 Fully                      i Inserted,    Full    Power,    Equilibrium    Xenon,            j 14000 mwd /T 4.3-11a                    Planar Average Power Distribution, Bank 3 and P2                    i Fully Inserted,      Full Power Equilibrium Xenon,                  )
14000 mwd /T 4.3-12                    Planar Average Power Distribution, PSCEA Banks                      ;
,                                Fully Inserted, Full Power, Equilibrium Xenon,                      I l
2000 mwd /T                                                          l l
Amendment B                  l l                                                  xvii                  March 31, 1980
 
L-CESSAR W h o.
                                                                                  ~!
LIST OF FIGURES (Cont'd)
CHAPTER 4                                    j risura  subioet w        4.3-13  Planar Average Power' Distribution, PSCEA Banks Fully Inserted, Full Power, Equilibrium Xenon, 9000 mwd /T                                                  '
              ~4.3-14  Planar Average Power . Distribution, PSCEA Banks w                  Fully Inserted, Full Power, Equilibrium Xenon, 14000 mwd /T 4.3-15  Planar Average Power Distribution, PSCEA Banks and      B-Bank 3 Fully Inserted, Full Power, Eglilibrium Xenon, 2000 mwd /T                                          -{
4.3-16  Planar Average Power Distribution, PSCEA Banks and Bank 3 Fully Inserted,    Full Power,  Equilibrium      -!
Xenon, 9000 mwd /T 4'.3-17 Planar Average Power Distribution, PSCEA Banks-and "e'
Bank 3 Fully Inserted, Full Power, Equilibrium Xenon, 14000 mwd /T' g      4.3-18  Unrodded Axial Power Distribution, BOC 4.3-19  Unrodded Axial Power Distribution at 50 mwd /T 4.3-20  Unrodded Axial Power Distribution at 4000 mwd /T i
4.3-21  Unrodded Axial Power Distribution at 9000 mwd /T 4.3-22  Unrodded Axial Power Distribution at 13000 mwd /T 4.3-23  Unrodded    Axial  Power Distribution,  EOC  (16000 mwd /MT) 4.3-24  Second Cycle Fuel Loading Pattern 4.3-25  Second Cycle Assembly Fuel Loadings, Waterhole and Shim Placements F
4.3-26  Assembly Average Burnup Distribution at Beginning of Second Cycle 4.3-27  Planar Average Power Distribution at Beginning of Second Cycle, Unrodded l
l
  ,                                                          Amendment F                l xviii              December 15, 1989
 
  '                                                                                                                                      i X
CESSARil!& c.
4
                                                                                                                                          \
LIL OF FIGURES (Cont'd)
CHAPTER 4 Figure                subient                                                                                                  {
4.3-28                Planar Average Power Distribution at Middle of                                                          ,
Second Cycle, Unrodded                                                                                .
4.3-29                Planar Average Power Distribution at End of Second Cycle, Unrodded                                                                                          !
4.3-30                Third Cycle Fuel-Loading Pattern l          4.3-31                Third Cycle Assembly Fuel Loadings, Waterhole and Shim Placement                                                                                  f 4.3-32                Assembly- Average Burnup                          Distribution    at              the Beginning of the Third Cycle l-        4.3-33                Planar Average Power Distribution at the Beginning
!                                of the Third Cycle, Unrodded                                                                            ,
4.3              Planar Average Power Distribution at the Middle of the Third Cycle, Unrodded 4.3-35                Planar Average Power Distribution at the End of the Third Cycle, Unrodded l
l          4.3-36                Daily        Reactor    Power                Maneuvering,    ASI                and                  i L                                Reactivity Control by'                        PSCEA    Only,  No . Boron, p                                100%-80%-100% Power Ramping Near BOC (0 hrs to 8 L
    -                            hrs into the maneuvering sequence) 4.3-36a              Daily        Reactor    Power                  Maneuvering,    ASI                and          g i                                Reactivity Control by                          PSCEA    Only,  No  Boron, l
100%-80%-100% Power Ramping Near BOC (8 hrs to 24' hrs into the maneuvering sequence)
          -4.3-37                Daily        Reactor    Power                  Maneuvering,    ASI                and Reactivity Control by                          PSCEA    Only,  No  Boron, 100%-80%-100% Power Ramping Near EOC (0 hrs to 8 hrs into the maneuvering sequence) l          4.3-37a              Daily        Reactor    Power                  Maneuvering,    ASI                and Reactivity Control by PSCEA Only,                              No  Boron, l
L                                100%-80%-100% Power Ramping Near EOC (8 hrs to 24 hrs into the maneuvering system)
Amendment F xix                                December 15, 1989 l
 
CESSAR !!nh.O.,
b LIST OF FIGURES (Cont'd)
CHAPTER d' Flaure    Subiect 4.3-38    [o vs Time for a 100%-50%-100% Load Following Transient 1
4.3-39  .
vs Time.for a 100%-50%-100% Load.Following            -{
Transient.                                                    i 4.3-40    Normalized Power Distribution of Unshimmed Assembly Used in Sample DNB Analysis in Section 4.4.2.2                i 4.3-41    Fuel- Temperature coefficient vs Effective Fuel Temperature                                                    '
i 4.3-42    Moderator Temperature      Coefficient > vs  Moderator Temperature at BOC 1 4.3-43    Moderator Temperature      Coefficient  vs  Moderator Temperature at EOC 1 4.3-44'    Moderator Density Coefficient vs Moderator Density                l 4.3-45    Fuel. Temperature Contribution to Power Coefficient              ,
at EOC 4.3-46    Control Element-Assembly Locations                        :
B 4.3-47    CEA Group Identification 4.3-48    Typical Power Dependent CEA Insertion Limit (PDIL)
                                                                                        =
4.3-49      Typical Integral Worth      vs  Withdrawal  at  Zero            4 Power, EOC 1 Conditions                                          j:
4.3-50      Typical Integral Worth vs Withdrawal at Hot Full Power, EOC 1 Equilibrium Xenon Conditions 4.3-51      Reactivity Difference Between Fundamental          and Excited States of a Bare Cylindrical Reactor 4.3-52      Expected Variation of the Azimuthal        Stability Index, Hot Full Power, No CEAs 4.3-53      PSCEA-Controlled      and    Uncontrolled      (Axial 0 Xenon-Induced Power) Oscillation
                                                                                          )
Amendment B                    1 xx                  March 31, 1988
 
Li
                                                                                        ')
C E S S A R K ncoil.
LIST OF FIGURES (Cont'd)                                  -
CHAPTER'4 Figure        Subiect                                                        -
4.3-541      Rod ' Shadowing Effect vs Rod Position for -Rod Insertion and Withdrawal ~ Transient at Palisades                ,
T 4.3-55      Typical Three'Sub-Channel Annealing 4.3-56        Geometry Layout                                                  ;
4.3-57        Comparison of Measured and Calculated            Shape L-                      Annealing Correlation for Palisades i
4.3-58        Typical      Temperature  Defect  vs  Reactor  Inlet        <
Temperature 4.3-59        Calculation-Measurement ITC Difference vs Soluble-Boron,.3-D ROCS (DIT) lJ                                                                                        '
4.3-60    . ROCS /DIT Reactivity from Core Follow Calculations, 14X14 Assembly Plants, Reload Cycles                            .
4.3-61        ROCS /DIT Reactivity from Core Follow' Calculations,          #
16X16 and 14X14-Assembly Plants 4.3-62        A    Divergent      Axial    (Xenon-Induced    Power)
Oscillation in an EOC Core with Reduced Power Feedback 4.3-63        Damping    Coefficient    vs  Reactivity    Difference Between Fundamental and Excited States                    B
          - 4.4-1        Core Wide Planar Power Distribution for Sample DNB Analysis 4.'4-2        Rod Radial Power Factors in Hot Assembly Quadrant for Sample DNB Analysis                                        ,
4.4-3        Typical Axial Power Distributions 4.4-4        Average Void Fractions and Qualities at the- Exit for Different Core Regions 4.4-5        Axial Distribution of Void Fraction and Quality in the Subchannel Adjacent to the Rod with the Minimum DNBR Amendment B xxi                March 31, 1988
 
o
                                                                              ~!
                .lCESSAR tiffn icm                                              ,
I l
t s
LIST OF FIGURES (Cont'd)
CEAPTER 4 Figure    subiect                                          -!
  '                                                                              l 4.4-6    ' Reactor Flow Paths                                <
                                                                              =
4.4-7    . Sensitivity of Minimum DNBR to Small Changes in    i Reactor Coolant Conditions l
4.4-8    Isometric View of the Reactor Coolant System 4
g:
l Iy i
+
4 xxii
 
CESSAR1!!L a 4
4.0              REACTOR 4.1             
 
==SUMMARY==
DESCRIPTION The reactor is of the pressurized water type using two reactor coolant loops.          A vertical. cross section of the reactor is shown in Figure 4.1-1.            The reactor core is composed of 241 fuel assemblies and 93 or more. control element assemblies (CEAs). The                          B fuel assemblies are arranged to approximate a right circular cylinder with an equivalent diameter of 143.6 inches- and an active length of 150 inches. The fuel assembly, which provides for 236 fuel rod positions (16 x 16 array), includes 5 guide tubes welded to spacer grids and is closed at the top and bottom i'              by and fittings.          The guide tubes each displace four fuel rod
!              positions and provide channels which guide the CEAs over their entire length of travel. In-core instrumentation is installed in                                                  ;
the central guide tube of selected fuel assemblies.                            The in-core
              ' instrumentation is routed into the bottom of the fuel assemblies through the bottom head of the reactor vessel.                          Figure 4.1-2 shows the reactor core cross section and dimensional relations l              between fuel assemblies, fuel rods and CEA guide tubes.
The fuel is low enrichment UO.3 in the form of ceramic pellets and is encapsulated in prepresstfrized Zircaloy tubes which form a hermetic enclosure.
The reactor coolant enters the inlet nozzles of the reactor vessel, flows . downward between the reactor vessel wall and the core barrel, and passes through the flow skirt section where the flow distribution is equalized, and into the-lower plenum. The coolant then flows upward through the core, removing heat from the fuel rods. The heated coolant enters the core outlet region where the coolant flows around the outside of control element assembly shroud tubes to the reactor vessel outlet nozzles.                            The control element assembly shroud tubes protect the individual neutron absorber elements of the CEAs from the effects of coolant cross flow above the core.
The reactor internals support and orient the fuel assemblies, control' element assemblies, and in-core instrumentation, and guide the reactor coolant through the reactor vessel. They also absorb static and dynamic loads and transmit the loads to the reactor vessel flange. They will safely perform their functions during normal operating, upset, and faulted conditions.                                The internals are designed to safely withstand forces due to dead weight, handling, temperature and pressure differentials, flow impingement, vibration, and seismic acceleration.                              All reactor components are considered Category I for seismic design.                                The design of the reactor internals limits deflection where                            required Amendment B 4.1-1                      March 31, 1988
 
t CESSAR tilLno.
by function. The stress values of all structural members under normal operating- and expected transient conditions are not greater than those established by Section III of the ASME Code.
The~effect of neutron irradiation on the materials concerned is t included- in the design evaluation. The effect of accident loadings on the internals is includes in the design analysis.
During normal operation, reactivity control is- provided by two    F independent' systems:  the Control Element Drive System and the Chemical- and Volume Control System. The J Control Element Drive
      ~ System controls - short tera reactivity changes and is used - for rapid shutdown. The Chemical and Volume Control System is used to compensate for- long-term reactivity changes and can make the reactor subcritical without the benefit of the Control Element Drive System. The design of the core and the Reactor Protective System prevents fuel damage limits from being exceeded for any single malfunction in either of the reactivity control systems.
During , accidents, the Safety Injection System also provides a safety-grade method of boron injection.                            F I
Amendment F 4.1-2              December 15, 1989
 
.rk;t 7
\
CEA EXTENSION SHAFT                                                                                                                                                  s
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                                            ,;                                                          ,-,,...,4      ,
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                                    \        l                                                                                  1 N                                                                                                    \
w y              :        mir i                gn    a    i
* REACTOR VESSEL I O I
ASSEMBLY FLOW SKIRT -        -
w7'                          '""' "
N                        _    _          ,        ,,        b              g;                      CORE STOP 1
N            1          )      '
                                                                                                              )
tY'%'
                                                                                                                            ,eIN CORE INSTRUMENTATION NOZZLE LOWER SUPPORT STRUCTURE Amendment F December 15,1989 m                                                                                                                                  Figure
        ;Jg                                                REACTOR VERTICAL ARRANGEMENT NA                                                                                      m:
 
(b FUEL ROD b
0.382" OD g    ,                                                            =-- 4.050" -*                                        -
[                      . GUIDE TUBE mennnnnMnanstmP)!)                                                          l ymmanhMMMMMMM l
n 1  ,                                                                ,
                                                                                                                                  ,2  i      ;,
i b-(7[
f"_',i                                                      -                        { ~_ ',)              i, C            0.124" p-                                                          -          .
g      6 d      ,
7.972"                      [                              ,_,                                      r ,,!                              .                  Si    i OUTSIDE                      ,:                                                                          :l        :                  I    i                  M k
FUEL RODS                    ::
                                                                        -                                      '1 L
1
:                    ''                    F      I n                                                                          1 l    l:                                            } l l 'p -                                                      ''
                                                                                                                                            $ j "_';                        (*{-f                i E-4 T , ','
t, is -
i,
                                                                                                                              -il t
                                                                                                                                                                                      -1      0.50s"
                                                              ,                                                                ,sr        i  t,  ,                                  ,g    FUEL ROD -
o                        ,9 t - sc~e-~scscscscs ~s N'                                                      l  5Hescifiesesngggmsd, PITCH 0.209" W AT E R ---*                  ~15 SPACES AT --*
GAP                                        0.506 = 7.59" REACTOR VESSEL 182*10
                                                                                                                                  /
CORE EQUlVALENT                                                                                            . ..
DI A.143.8*
                                                                      &          /                                                                                      '
7
                                                                    /
                                  /
:                                                                                                                                        /
                                  /                ,                                                                                                                                  l
                                                                                                                                                                                              /
                                                                                                                                                  -                                            Amendment F CORE SUPPORT                                                                                                                                                    December 15,1989 BARREL 157" 10                                                                      / /
REACTOR CORE CROSS SECTION
                -gj                                                                                                              241 FUEL ASSEMBLIES                                                  4.12
 
CESSAR Manio,.
4.?        FUEL SYSTEM DESIGN 4.2.1        DESIGN BASES 4.2.1.1        Fuel Assembly The fuel assemblies:are required to meet design criteria for each design condition listed below - to - assure that the functional requirements are met.      Except where specifically noted, the design bases presented in this section are consistent with those used for previous designs.-
A. Nonoperation and Normal operation    (Condition I)
Condition I    situations are those which are planned or expected to occur in the course of handling,            initial shipping, storage, reactor servicing and power operation (including    maneuvering of the plant).          Condition  I situations must be accommodated without fuel assembly failure and without any effect which would lead to a restriction on subsequent operation of the fuel assembly.
The guidelines stated below arp used to determine loads during condition I situations:
: 1. Handling and Fresh Fuel Shipping Loads correspond to the maximum possible axial and lateral loads and accelerations imposed on the fuel assembly by shipping and handling equipment during these periods, assuming that there is no abnormal contact between the fuel assembly and any surface, nor any equipment malfunction.      Irradiation effects on material properties are considered when analyzing the effects of handling loads which occur during refueling.
Additional information regarding shipping and handling loads is contained in Section 4.2.3.1.5.
: 2. Storage Loads on both new and irradiated fuel assemblies reflect storage conditions of temperature, chemistry, means of support and duration of storage.
: 3. Reactor Servicing Loads on the fuel assembly reflect those encountered during refueling, inspection, and reconstitution.          F l
Amendment F 4.2-1                December 15, 1989
 
CESSAR1Pe h on 4 '. Power Operation Loads - are ' derived from conditions encountered during transient and steady-state operation in the design power range. .(Hot operational testing, system startup, hot standby,    operator-controlled transients within specified rate limits and system shutdown are included in this category.)
: 5. Reactor Trip Loads correspond to those produced in the fuel assembly by. control  element    assembly  (CEA)    motion  and deceleration.
B. Upset condition (Condition II)
Condition II situations are unplanned events and operating basis earthquakes (OBEs) which may occur with moderate frequency during the life of the plant. The fuel assembly
            -design should have : the capability to withstand any upset condition with margin to _ mechanical failure and with no permanent effects which would prevent continued normal operation. Events classified as Upset Conditions are listed below:
: 1. Operating Basis Earthquake (OBE)-
T-    2. Uncontrolled CEA Withdrawal
: 3. Uncontrolled Boron Dilution
: 4. Partial Loss-of-Coolant Flow
: 5. Idle  Loop Startup      (in  violation  of  established operating procedures)
: 6. Loss of Load (reactor-turbine load mismatch)
: 7. Loss of Normal Feedwater                                      B
: 8. Loss of Offsite Power 9.-  Excessive Heat Remcval (feedwater system malfunction)
: 10. CEA Drop
: 11. Accidental Depressurization      of  the  Reactor  Coolant System (RCS)
C. Emergency Conditions (Condition III)
From Chanter 15 and Minor Fuel Handline Accidents Condition III events are unplanned incidents and minor fuel handling accidents which might occur infrequently during plant life. Rod mechanical failure must be prevented for any Condition III event in any area not subject to extreme Amendment B 4.2-2                March 31, 1988
 
CESSAR !!!O,ce g
J l
local. conditions (e.g.,                  in any rod not immediately adjacent                                        J to the impact surface during a fuel handling accident).
Incidents          classified an. Emergency Conditions are . listed-                                                  ]
below:
: 1. Complete loss or . interruption of primary coolant- flow at 100% power, excluding the occurrence of a reactor coolant pump locked rotor
: 2. Steam bypass malfunction                                                                  B
: 3. Minor. fuel handling accident (fuel assembly and grapple remain connected)
: 4. Inadvertent. loading of a fuel assembly into improper position.
D.      Faulted Conditions (Condition.IV)
                                                                                                                                =
Condition IV incidents - are postulated events (as discussed                                                  E-in Chapter 15)                    and the Safe Shutdown Earthquake (SSE), LOCA                                  .
(Mechanical Excitation only), combined SSE and LOCA, and                                                          '
major fuel handling accidents whose consequences are such that integrity and operability of the nuclear energy system may-be impaired.                      Mechanical fuel failures are permitted, but they must not -impair the operation of the Engineered Safety Features (E5F) systems to mitigate the consequences of the postulated event.                        Events classified as Faulted Conditions are listad below:                                                                                  ; '
l F                    l
: 1. Safe Shutdown Earthquake (SSE)                                                                                  i
: 2. Loss-of-Coolant Accident (LOCA)
: 3. Locked Reactor Coolant Pump Rotor
: 4. Major Secondary System Pipe Rupture                                                        B
: 5. CEA Ejection
: 6. Major Fuel Handling Accident (fuel assembly and grapple                                                          j are disengaged).                                                                                                '
4.2.1.1.1                    Fuel Assembly 8tructural Integrity criteria l
For each of the design conditions, there are structural criteria which apply to the fuel assembly and its components, but not to individual fuel rods.                        These criteria are listed below and give the allowable stresses and functional requirements for each design condition.                          Criteria  for individual          fuel  rods          are discussed separately in Section 4.2.1.2.
Amendment F 4.2-3                      December 15, 1989
  . . _ _ _ . . .          . _ . _ _ _ _ _ _ _ . _          _    _  _  __  _ _ _ . _ _        _ _ . .._ _          ~ _ - --
 
v.
(:
* s CESSAR tinh                              .
s o
A.            Design Conditions I and II I
P, 5 S, P, + Pb $ F,S, Under~ cyclic loading conditions, stresses must be such that                                              !
the cumulative' fatigue damage factor does not exceed 0.8.                                                j The cumulative fatigue damage factor is defined, as the sum of the ration of the number of cycles at a given cyclic stress (or strain) condition to the maximum number permitted                                              ,
for that condition.            The selected limit of 0.8 is used in place of 1.0 (which would correspond.to the absolute maximum fatigue . damage            factor permitted)        to  provide                additional.
margin in the design.                                                                                      j During the OBE, fuel assembly deflections must be such that ~
permanent deformations are limited to a value allowing the                                                {
CEAs to scram.
B.          Design Condition'III P,5 1.5 S,                                                                                                l P, + Pb $ 1.5 F,S,,
C.          Design Condition IV                                                                                        !
1 P, $ Sy                                                                                                    i P, + P3 $ F,Sy whereSg=smallervalueof2.4S,or0.7Su*
l
: 1. If the equivalent diameter pipe break in the LOCA does not            exceed 0.5 square feet,            the          fuel        assembly deformation shall be limited to a value not exceeding                                                i the deformation which would preclude                                    satisfactory insertion of the CEAs.
: 2. For pipe break sizes greater than 0.5 square feet, deformation of structural components is limited to-maintain the fuel in a coolable array. CEA insertion is not required for these events as the appropriate safety analyses do not take credit for CEA insertion.
1 4.2-4
 
o CESSAR %HMiean.=
3 . .-  For the upper and fitting holddown springs,. calculated.
                  ' shear stress must not exceed the minimum yield stress in shear.
: 4.      For the spacer grids, the predicted impact loads.must be less than the tested grid capability, as defined in Reference 50..
: 8.      During.the SSE,. fuel assembly deflections must be such that permanent ~ deformations are limited to a value allowing the CEAs to scram.
D.-  Nomenclature The symbols used in defining the allowable stress levels are as follows:
P,= Calculated general primary membrane stressI ")                  u Pb = Calculated primary bending stress S, = Design stress ' intensity value as defing) by Section III, ASME Boiler and Pressure Vessel Code S    = Minimum unirradiated ultimate tensile strength u
factor corresp            to tne particular cross F, = Shapesection being analyzedging S4=Designstressintensityvalueforfaultedconditions ThedefinitionofSyasthelesservalueof2.4S                    and 0.7 S is
    . contained in        the  ASME    Boiler    and' Pressure" Vessel  C8de, Section III.
(a)  P    and P are defined by Section III,          ASME Boiler and PEessure Yessel Code.
(b)  With the exception of zirconium base alloys, the design stress intensity values, S          of materials not tabulated by the Code are determined in T,he same manner as the Code.        The design stress intensity of - zirconium base alloys shall not exceed two-thirds of the unirradiated minimum yield strength at temperature.      Basing the design stress intensity on the unirradiated yield strength is conservative because the yield strength of zircaloy increases with irradiation.          The use of the two-thirds factor ensures 50% margin to component yielding in response to primary stresses. This 50% margin
: 4. 2 -5
 
s                                                                                  ,
[
CESSAR Minein.
i together with its application to the minimum unirradiated properties and the general conservatism applied in the establishment of design conditions is sufficient to ensure an adequate design.
(c)  The shape    factor,  F    is defined as the ratio of- the
-                  " plastic" moment - (all", fibers just at the yield stress)  to the initial yield amount (extreme fiber at the yield stress        i and all other fibers stressed in proportion to their                i
      ,  .      distance from the neutral axis). The capability of cross sections loaded in bending to sustain moments considerably        :
in excess of that required to yield the outermost fibers is        .
discussed in Timoshenko (see Reference 1).
4.2.1.1.2        Material Selection The fuel assembly grid cage structure consists of 10 Zircaloy-4 spacer grids, 1 Inconel 625 spacer grid (at the lower end) , 5 Zircaloy-4 guide tubes, 2 stainless steel end fittings, and 4              ,
            - Inconel X-750 coil springs.      Zircaloy-4, selected for fuel rod cladding, guide tubes and spacer grids, has a low neutron absorption
* cross section, and high corrosion resistance to the reactor water environment. Also there is little reaction between the cladding and fuel or fission products.          As described in Section 4.2.3, Zircaloy-4 has demonstrated. its ability as a cladding, CEA guide tube, and spacer grid material.
The bottom spac'er grid is of Inconel- 625 and is welded to the
            -lower and fitting.      In this region of local inlet turbulence, Inconal 625' was selected rather than Zircaloy-4 to provide additional strength and relaxation resistance.      Inconel 625 is a very strong- material with good ductility, corrosion resistance and stability under irradiation at temperatures below 1000*F.
The fuel assembly upper and lower end fitting are of cast 304 stainless steel and the upper and lower and fitting posts are Type 304 stainless steel machined components. This material was selected based on considerations of adequate strength and high corrosion resistance. Also, Type 304 stainless steel has been used successfully in almost all pressurized water reactor environments, including all currently operating C-E reactors.
4.2.1.1.3        Control Element Assembly Guide Tubes All CEA guide tubes are manufactured in accordance with ASTM B353, Wrought Zirconium and Zirconium Alloy Seamless and Welded Tubes for Nuclear Service, with the following exceptions and/or additions:
4.2-6
 
=                                                                                                                                        i
:CESSAR1!Mieu.
A. Chemical Properties' Additional limits are placed on oxygen, carbon and silicon, B. . Mechanical Properties Minimum values are specified for the-tensile-strength, yield                                              B strength and total elongation at room temperature and high                                                          i temperature.
C. Dimensional Requirements-Permissible Tolerance-Dimension                                                        fin.)
OD                                                          10.003 ID                                                          0.005 4.2.1.1.4        Sircaloy-4 Bar stock                                                                            ;.
Zircaloy-4 bar stock is fabricated in accordance with Grade                                                    B R60804, ASTM B351, Hot-Rolled and Cold-Finished Zirconium and Zirconium Alloy Bars, Rod and Wire for Nuclear Application, with-the following exceptions and/or additions:
A. Chemical Properties 1
Additional limits are placed on oxygen, carbon and silicon                                                B  _
content.
B. Metallurgical Properties The maximum average grain size is restricted.
4.2.1.1.5        Sirealcy-4 Strip Stock All-Zircaloy-4 strip stock is fabricated in accordance with Grade                                              8 R60804, ASTM B352, Zirconium and Zirconium Alloy Sheet, Strip and Plate for Nuclear Application,                with the following exceptions and/or additions:
A. Chemical Properties Additional limits are placed on oxygen, carbon and silicon                                                B content.
B. Metallurgical Properties The maximum average grain size is restricted.
Amendment B 4.2-7                                March 31, 1988
 
l k !!!hkh$k!I SER$ftCATION-C.-        . Mechanical Properties o
Each sample shall be-tested for hardness in accordance with                                                      !
the procedura described in ASTM E18 (Standard Test Method
: i.                    for Rockwell Superficial Hardness of Metallic Materials) .                                              B L                    The Rockwell hardness' is limited to a value to ensure adequate; material ductility.
l D.          Coefficient of Thermal Expansion                                                                                l l
Axial direction - see Reference 2 l
E.          Irradiation Properties The yield and tensile strengths are enhanced by irradiation.                                                    i
!                    The  stress      relaxation with ~ irradiation                        at-              operating              i temperatures proceeds at a rapid rate until nearly complete.                                                    1 The irradiation induced growth is documented in References 3-                                                    l and 4.
4.2.1.1.6                  Stainlees steel. castings.                        ,
;    4  All stainless steel castings are fabricated in accordance with l        Grade CF-8, ASTM A744, with the.following. addition:
L                                                                                                                            B Heat treatment is specified to meet designated cooling rate and the acceptable level delta ferrite.
I,        4.2.1.1.7                  Stainless Steel Tubing L        All stainless steel tubing is fabricated in accordance with ASTM                                                            I A269, with the following addition:
Carbon content is limited on tubing to be welded.
!        4.2.1.1.8                  Inconel X-750 compression springs l        All Inconel springs are fabricated in accordance with AMS 5699C.
4.2.1.1.9                  Inconel 625 Bottom Spacer Grid Strip Material l
Inconel spacer grid strip naterial is procured in accordance with the specification plate,        sheet, and  for nickel-chromium-molybdenum-columbium strip, Specification ASTM B433, withalloy                                  the ,l following additions:
A.        Check analysis is required, and B.        Material is required to pass a specified bend test and hardness requirement.
Amendment B 4.2-8                          March 31, 1988
        .  . - _ _ .          __ _ _        - _ _ ~ _ _      _ _ . . _ _ _ _ _            . _ . . _ _ _ _ . . _ . .        - _ , . _
 
CESSARiinL mn 4.2.1.2        Fuel Rod 4.2.1.2.1        Fuel Cladding Design Limits The fuel cladding is designed to sustain the effects of steady-state and expected transient operating conditions without exceeding acceptable levels of stress and strain. Except where specifically noted, the design bases presented in this section are consistent with those used for previous core designs. The fuel rod design accounts for cladding irradiation growth,-
external pressure, differential expansion of fuel and clad, fuel swelling, densification, clad creep,                              fission and other gas releases,    initial- internal helium pressure,- thermal stress, pressure. and temperature _ cycling, and flow induced vibrations.
The structural criteria discussed below are- based on the following    for  the _ normal,                            upset, and emergency    loading combinations identified in Section 4.2.1.1. For a discussion of                            _.
the thermal / hydraulic criteria, see Section 4.4.1.
A. During normal operating and _ upset conditions, the maximum primary tensile stress in the Zircaloy clad shall not exceed two-thirds of the minimum unirradiated yield strength of the material at the applicable temperature.                -
The corresponding
                -limit under emergency conditions is the material yield strength. The use of the unirradiated material yield strength as the basis for allowable- stress is conservative because the yield. strength of Zircaloy increases with irradiation. The use of the two-thirds factor ensures 50%
margin to component yielding in response to primary stresses.- This 50% margin, together with its application to the  minimum unirradiated properties and the general conservatism applied in the establishment of design conditions, is sufficient to ensure an adequate design.
B. Net unrecoverable circumferential strain shall not exceed 1%
as predicted by computations considering clad creep and fuel-clad interaction effects.
Data from O'Donnell (Reference 5) and Weber (Reference 6) were  used to determine the present 1% strain limit.
O'Donnell developed an analytical failure curve for Zircaloy cladding based upon the maximum strain of the material at its point of plastic instability.                            O'Donnell compared his analytical curve to circumferential strain data obtained on irradiated coextruded Zr-U metal fuel rods tested by Weber.
The correlation was good, thus substantiating O'Donnell's instability theory. Since O'Donnell performed his analysis, additional data have been derived at Bettis (References 7-9) and AECL (References 10 and 11).
4.2-9
 
s            >                .
U .N "CESSAR1RWie rion TheseL new- data    are shown in Figure 4.2-1, along with-O'Donnell's curve and Weber's data.      This curve was then adjusted because of differences in anisotropy, stress states and strain, rates; and the design limit was set at 1%.
The conservatism of the clad strain calculations is provided
                    .by: the selection of adverse initial conditions and material behavior: assumptions, and by the assumed operating history.
The = acceptability of the 1%' unrecoverable circumferential-strain' limit is- demonstrated    by data from irradiated Zircaloy-clad fuel rods which show no cladding failures (due to strain) at or below this level, as illustrated in Figure
!            i 4.2-1.-
C. The clad will be initially pressurized with helium .to an amount sufficient to prevent gross clad deformation under the combined ' etfacts of external pressure and long-term n              creep. The clad design will not rely on the support of fuel pellets or the holddown spring to prevent gross deformation.
D. Cumulative- strain cycling usage, defined as the sum of the ratios. of the. number of cycles in a given effective strain range - (as ) to the permitted number (N) at that range, as taken from Figure 4.2-2, will not exceed 0.8.
_ The cyclic strain limit design curve shown on Figure 4.2-2 is based upon the Method of Universal Slopes developed by S. S. Manson (Reference 12) and has been adjusted to provide a strain cycle margin for the effects of uncertainty and irradiation _  The resulting curve has been compared with known data on the cyclic loading of Zircaloy. and has been shown to be conservative. Specifically, it encompasses all the data of O'Donnell and Langer (Reference 13).
As discussed in Section 4.2.1.2.5, the fatigue calculation method includes the effect of clad creep to reduce the pellet-to-clad    diametral- gap during that portion of operation when the pellet and clad are not in contact. The same model' is used for predicting clad fatigue as is used for predicting clad strain. Therefore, the effects of creep and fatigue loadings are considered together in determining end-of-life clad strain.      Moreover, the current fatigue damage calculation method includes a factor of 2 which is applied to the calculated strain before determining the allowable number of cycles associated with that strain.
This, in combination with the allowable fatigue usage factor 0.9, ensures a considerable degree of conservatism        (see Figure 4.2-2).
1 l
u Y                                                4.2-10
 
1 CESSARRBLn..
1 1
1 E.        There is no specific limit on lateral fuel rod deflection                                                    {
for structural integrity considerations except that which is                                                ,
brought      about through application of cladding stress criteria. The absence of a specific limit on rod deflection                                                  i is justified because it is the fuel assembly structure, not the individual fuel rod, that is the limiting factor for fuel assembly lateral deflection.                                                                            .
i F.          Fuel rod internal pressure increases with increasing burnupt                                                !
toward end-of-life, the total internal pressure, due to the                                                  !
combined effects of the initial helium till gas and the                                                      :
released' fission gas, can approach values comparable to the                                                -
external coolant pressure.              The maximum pred),cted fuel rod internal pressure will be consistent with the following                                                      '
criteria.
: 1. The primary stress              in the cladding resulting                                      from  I differential pressure will not exceed the stress limits specified earlier in this section.                                                                  f ,
i
: 2. The internal pressure will not cause the clad to creep outward from the fuel pellet surface while operating at                                              ;
the design ; peak linear heat rate for normal operation.
In determin:ng compliance with this criterion, internal pressure is calculated for the peak power rod in the                                                  ,
reactor, including accounting for the maximum computed                                                ,
fission gas release.            In addition, the pellet swelling rate (to which the calculated clad creep rate is compared) is based on the observed swelling rate of                                                _
                      " restrained" pellets - (i.e. , pellets in contact with clad), rather than on the greater observed swelling                                                  i behavior of pellets which are free to expand.
The criteria discussed above do not limit fuel rod internal pressure to values- less than the primary coolant pressure,            and the occurrence of positive differential    pressures            would not adversely affect normal operation if appropriate criteria for cladding                                                ,
stress, strain, and strain rate were satisfied.
The design limits of the fuel rod cladding, with respect to G.
vibration considerations, are incorporated within the fuel assembly design.      It is a requirement that the spacer grid                                              -
intervals, in conjunction with the fuol rod stiffness, be such that fuel rod vibration, as a result of mechanical or flow induced excitation, does not result in excessive wear of the fuel rod cladding at the spacer grid contact areas.
E.
4.2-11
      - _ _ _ _ _ ,                  _      = - - =    -
 
i i
  . CESSAR RHMen..                                                                                                  !
4.2.1.2.3        Puel Rod Cladelat Properties 4.2.1.3.2.1        Meehanical Properties                                                                        :
A. Modulus of Elasticity i
Young's Modulus is as specified in Reference 14.
B. Poisson's Ratio i
Poisson's Ratio is as specified in Reference 14.                                                          l
. C. Thermal coefficient of Expansion i
I                                                                                                                    '
Diametral direction Thermal Coefficient of Expansion is specified in Reference 14.
D. Yield Strength                                                                                            -
Yield strength in as specified in Reference 14.
The cladding stress limits identified in Section 4.2.1.2.1
* are based on values taken from the minimum yield strength curve at the appropriate temperaturas.                                        The limits are              .
applied over the entire fuel lifetime, during conditions of reactor heatup and cooldown, steady state operation, and normal power cycling.                      Under these conditions, cladding temperatures and fagg fluences can range from 70 to 750'F and from 0 to 1 x 10            nyt, respectively.
E. Ultimate Strength Ultimate tensile strength is as specified in Reference 14.
F. Uniform Tensile Strain Uniform tensile strain is as specified in Reference 14.
Uniform approaches tensile 1%  at strain 6 x              10 g            the nyt and irradiated remains condition relatively          ,
constant (Section 4.2.1.2.1).
G. Hydrostatic Burst Test The cladding specification requires that two samples from each lot of cladding be subjected to room                                            temperature hydrostatic  burst        tests.                        To    be  acceptable,      the burst pressure must exceed a minimum value, based on the cladding geometry  and    specified                          tensile        properties,      and  the circumferential elongation must exceed a prescribed minimum value.
4.2-12
 
l CESSAR tinh..                                                                                                                ;
l 4.2.1.2.2.2                                Dimenaional Requirements i
A.                Tube straightness is limited to 0.010 in./tt, and inside diameter and wall thickness are tightly controlled,                                                        i l
: 3.                Ovality in measured as the dif ference between maximum and                                                i minimum inside diameters and is acceptable if within the                                                  i diU4 ster tolerances.                                                                                      j J
l-    c.                Outside diameter is specified as 0.382 1 0.002 inches.                                                    ]
I D.                Inside diameter is specified as 0.332 1 0.0015 inches.                                                    ;
1 E.                Eccentricity is defined as the difference between maximum                                                !
and minimum wall thickness at a cross-section, and is                                                    ,
specified as 0.004 inches maximum.
F.                Wall thickness is specified as 0.023 inches minimum (the                                                  '
nomina 3 value reported elsewhere is based on the nominal CD and ID).                                                                                                ,
4.2.1.2.2.2                                Natallurgical Properties A.                  Hydride Orientation                                                                              ,
A restriction is placed on the hydride orientation factor                                        i      '
for any third wall thickness of the tube cross-section                                  B (inside,          middle,      or outside).      The hydride orientation                    "
factor,            defined as the ratio of the number of radially                                        ,
oriented hydride platelets to the total number of hydride platelets, shall not excsed 0.3. The independent evaluation of three portions cf the cross-section is included to allow                                            i for the possibility that hydride orientation may not be                                                r uniform across the entire cross-section.
4 2.1.2.2.4                                Chemical Prop 6rties All fuel rod cladding is manufactured in accordance with ASTM B353, Wrought Zirconium and Zirconium Alley Seamless and Welded Tubes for Nuclanr Service, except additional limits are placed on B
oxygen, silicon, and carbon content.
4.2.1.2.3                                Fuel Rod Component Propertius 4.2.1.2.3.1                                sircaloy-4 Bar stock All Zircaloy-4 bar stock is fabricated in accordance with ASTM B351, Hot-Rolled and Cold-Finished Zirconium and Zirconiun Alley Bars, Rod and Wire for Nuclear Application, with the following exceptions and/or additiont, Amendment B 4.2-13                    March 31, 1980
 
1 l
CESSARRHhr.                                                                                            !
A.      Chemical Properties Additional limits are placed on oxygen, carbon and silicon contant,                                                                                      '
B. Metallurgical Properties                                                                      :
The maximum average grain size is restricted.
C.      Non-Destructive Testing Ultrasonic inspection is required.
4.k.1.2.3.2                  stainless steel Compressica springs All stainless steel springs are fabricated in accordance with AMS                                    .
5688, Revision G.
* p
,  4.2.1.2.4                  00 2 Fuel Pellet Properties                                              i 4.2.1.a.4.1                  chemical. composition salient      points          regarding the structure, composition, and propertiar of the UO fuel pellets are discussed in the following subsections. Where $he effect of irradiation on a specific item is  considered to be of sufficient importance to warrant                                            '
reflection in the design or analyses, that effect is also discussed.
A.      Chemical          analyses    are      performed                  for the  following constituents:
: 1. Total Uranium
: 2. Carbon                                                                                'i
: 3. Nitrogen                                                                              -
4.-  Fluorine
: 5. Chlorine and Fluorine
: 6. Iron
: 7. Thorium
: 8. Nickel
: 9. Calcium and Magnesium                                                                  .
: 10. Chromium                                                                            B
: 31. Aluminum
: 12. Silicon Amendment F 4.2-14                                December 15, 1989
                            ~            -
 
t CESSAR titMem..
I B.                              The oxygen-to-uranium ratio is maintained between 1.99 and                                                          g      i 2.02.                                                                                                                      ,
C.                              The sum of the cross-sections of the following impurities shall not exceed a specified equivalent thennal-neutron capture cross-section of natural boron:
: 1.        Boron.
: 2.        Silver                                                                                                          ,
: 3.        Cadmium                                                                                                        '
: 4.        Gadolinium
: 5.        Europium
: 6.        Samarium                                                                                                        ;
: 7.        Dysprosium D.                              The total hydrogen content of finished ground palla':s is restricted.
E.                              The nominal enrichment of the fuel pellets will be specified and.shall be held within 10.05 wt% U-235.
4.2.1.2.4.2                                        Microstructure A.                              The pellet fabrication process will maximize the pore content of pellets in a specified range.                                                            Acceptable porosity distribution will be determined by comparison of                                                                  i approved visual standards with photo-micrographs from each                                                              -
pellet lot.
B.                              The average grain size shall exceed a specified minimum                                                                    '
size.                                                                                                                    ;
4.2.1.2.4.3                                        Density A.                              The density of the sintered pellet after grinding shall be between 93.5 and 96.0% of theoretical d sity (TD), based on a U0 theoretical density of 10.96 g/cm 2
B.                              The in-pile stability of the fuel is ensured by the use of an NRC-approved out-of-pile test during production.                                                            The        l details of this test, and the associated rationale, are presented in Reference 14.
C.                              The effects of irradiation on the density of sintered UO pellets        are      treated    in  compliance                with                    the    intent      oE y      .
Regulatory Guide 1.126, through use of the NRC-approved model for fuel evaluation presented in References 15-17.
Amendment F 4.2-15                                                December 15, 1989
_ _ - _ = - . . _ _ _ . _ _ _ _ . - _ _ _ _ _ _ _ _ _ _
                                                                    ...,e..w e ._w.,        - ,.,,w....,,,,,,      .v.,,,,,.,-ym.--,,-.-%-            , _ , -
 
I CESSAR REMean .
4.2.1.2.4.4        Thermal Properties A. Thermal Expansion
      .The' thermal expansion of UO., is described by the following temperature dependent equatibns (References 18 and 19);
      % Linear Expansion = (-1.723 x 10-2) + (6.797 x 10~4 T)
                                + (2.896 x 10 ~7T2) from 25'c to 2200*C; and,                                                                      ,
      % Linear Expansion = 0.204 + (3 x 10 ~4 T)                      + (2 x 10 ~7T2)                ;
                                + (10- OT3) above 2200*C, where T is the temperature in degrees Celfius.
B. Thermal Emissivity A value of 0.85 is used for the thermal emissivity of Uo                                          l pellets.over the temperature range 800 to 2600'X (Referencek                                      !
20-22).                                                                                            l C. Melting Point and Thermal Conductivity The variation of melting point and thermal conductivity with burnup is discussed in References 15-17.
L D. Specific Heat of Uo 2
The specific heat of            UO.,  is described by the following                            ,
temperature dependent equations (Reference 23)
For T < 2240'F:
6
                                        -3                .4        x 0 C  = 49.67 + 2.2784 x 10          T-                              ; and, P
(T + 460)2 l
For T h 2240'F:
                                                                ~4                      -8 3 C = -126.07 + 0.2621T - 1.399 x 10                          T + 3.1786 x 10        T p
          - 2.483 x 10-12T4, 4.2-16
 
I CESSARtlRLn..                                                                                                                                                                                      I l
l i
h where:
3 C  = specific heat, BTU /ft                                  *F; and, p
T  = temperature, 'F.
4.2.1.2.4.5                    Mechanical Properties A. Young's Modulus of Elasticity The static modulus of elasticity of unirradiated fpel of 97%                                                                                                                                  !
TD and deformed under a strain rate of 0.097 hr" is given by Reforence 24:
6                                                                                                                                      '
E = 14.22 (1.6715 x 10                                  - 924.4T),
where E = modulus of elasticity in psis and,                                                                                                                                                        !
T = temperature in 'C in the range of 1000 to 1700*C.
B. Poisson's Ratio i
The Poisson's Ratio of polycrystalline UO, has a value of                                                                                                                                    -
0.32 at 25'C based on Reference 66. Thn same reference notes a 10% decrease in value over the range of 25 to
,      1800'C. Assuming the decrease is linear, the temperature dependence of the Poisson's Ratio is given by:
                                                          -5 (T-25),
v = 0.32 - 1.8 x 10 where                                                                                                                                                                                        -
l      v = Poisson's Ratio T = temperature in 'C in the range of 25 to 1800'C.
At temperatures above 1800'C, a constant value of 0.29 is
,      used for Poisson's Ratio.                                                                                                                                                                    ,
4.2.1.2.5            Fuel Rod Pressurisation Fuel rods are initially pressurized with helium for two reasons:
A. To preclude clad collapse during the design life of the fuel. The internal pressurization, by reducing stresses i      from differential pressure, extends the time required to produce creep collapse beyond the required service life of the fuel; and, i
l 4.2-17 i
 
CESSAR IMnem..
B. To improve the thermal conductivity of the pellet-to-clad gap within the fuel rod. Helium has a higher coefficient of thermal conductivity than the gaseous fission products.
In unpressurized fuel, the initially good helium conductivity is eventualJy degraded through the addition of the fission product gases    released                    from          the      pellets.      The initial  helium pressurization results in a high helium to fission products ratio over the design life of the fuel with a corresponding increase in the gap conductivity and heat transfer. The effect of fuel rod power level and pin burnup on fuel rod internal pressure has been studied parametrically.
The initial helium fill pressure will be 380 psig. This initial fill pressure will be sufficient to prevent clad collapse discussed in Section 4.2.3.2.7 and will pro 6uce a maximum EOL                                  '
internal pressure consistent with the criteria of Section 4.2.1.2.1.      The calculational methods employed to generate internal pressure histories are discussed in References 15-17.                                  ,
4.2.1.2.5.1                      Capacity for Fission Gas Inventory The greater portion of the gaseous fission products remain either within the lattice or the microporosity of the UO fuel pellets and do not contribute to the fuel rod intedal pressure.
However, a fraction of the fission gas is released from the pellets      by diffusion and pore migration and thereafter contributes to the internal pressure.
1 The determination of the effect of fission gas generated in and released from the pellet column is discussed                                        in Section 4.2.3.2.2. The rod pressure increase which results from the release of a given quantity of gas from the fuel pellets depends upon the amount of open void volume available within the fuel rod and the temperatures associated with the various void volumes.
In the fuel rod design, the void volumes considered in computing internal pressure are:
o  Fuel rod upper end plenum; o  Fuel-clad annulus; o  Fuel pellet-end dishes and chamfers; and, o  Fuel pellet open porosity.
These volumes are not constant during the life of the fuel. The model used for computing the available volume as a function of burnup and power level accounts for the effects of fuel and clad thermal expansion, fuel pellet densification, clad creep, clad growth, and irradiation induced swelling of the fuel pellets.
4.2-18
 
CESSAR !!Wcar                                                                  l i
l l
4.2.1.2.5.2        Fuel Rod Plenum Design The fuel rod upper and plenum is required to serve the following functions:
A. Provide    space for axial thermal          expansion  and    burnup    j
  .      swelling of the pellet columns l  B. Contain the pellet column holddown springi and,                          l l
C. Act as a plenum region to ensure an acceptable range of fuel              )
rod internal pressure.                                                    ;
l  Of these functions, item C is expected to be the most limiting                .
;  constraint on planum length selection, since the range of                    ;
:  temperatures in the fuel rod, together with the effects of                    I swelling, thermal expansion, and fission gas release, produce a                l wide range of internal pressure during the life of the fuel. The fuel    rod  plenum pressure will be            consistent  with    the pressurization and clad collapse criteria specified in Section 4.2.1.2.1.
4.2.1.2.5.3        Outline of Procedure Used to Sise the Fuel Rod Plenum A. A parametric study of the effects of plenum length on maximum and minimum rod internal pressure is performed.
Because the criteria pertaining to maximum and minimum rod            - !
internal pressure differ, the study is divided into two sections:
: 1. Maximum Internal Pressure Calculation                            ,
Maximum rod pressure is limited by the criteria as specified in Section 4.2.1.2.1.F.        Maximum and-of-life pressure is determined for each plenum length by including the fission gas released, selecting conservative values for component dimensions and properties, and accounting for burnup effects            l on component dimensions.            The primary cladding stress produced by each maximum pressure is then compared to the stress limits to find the margin available with each plenum length. Stress limits are listed in Section 4.2.1.2.1.
: 2. Minimum Internal Pressure / Collapse Calculation Minimum rod pressure is limited by the criterion that no rod will be subject to collapse during the design lifetime. The minimum pressure history for            each plenum length        is determined by neglecting fission gas release, selecting a conservative    combination of component dimensions and properties, and accounting for dimensional changes during 4.2-19
 
i CESSAR WWeno                                                                                                                                          !
l I
irradiation,                including the effects of cladding creep,                                                                            ,
cladding growth, pellet densification, pellet swelling, and                                                                                        1 thermal expansion.                        Each minimum pressure history is input to the cladding collapse model (Reference 25) to establish the acceptability of the associated plenum length,                                                                                                i B.- For each plenum length,                          there is a resultant range of                                                                  I acceptable initial fill pressures.                                                                                The optimum plenum              J 1ength is generally considered to be the shortest which                                                                                          :
satisfies all criteria related to maximum and minimum rod                                                                                          l internal              pressure            including                            a                    range          sufficient      to        :
accommodate a reasonable manufacturing tolerance on initial                                                                                        l fill pressure.
C. Additional information on those factors which have a bearing on determination of the plenum length are discussed below:
: 1.            Creep    and dimensional stability of the fuel rod                                                                                ;
7                                    assembly itifluence the fission gas release model and                                                                              '
internal pressure calculations, and are accounted for-in the procedure of sizing the fuel rod plenum length.                                                                              i Creep in the cladding is accounted for in a change in clad inside diameter, which in turn influences the                                                                                  )
fuel / clad gap.              The        gap change varies the gap                                                                1 conductance in the FATES computer code (References                                                                                  i 15-17) with resulting change in annulus temperature, internal      pressure,          and fission gas release.                                                            In addition, the change in clad inside diameter causes a change in the internal volume, with its resulting                                                                                  i effect on          temperature                        and pressure.                                            Dimensional stability considerations affect the internal volume of the fuel rod, causing changes in internal pressure and temperature.              Fuel pellet densification reduces the                                                                    ,
stack height and pellet diameter.                                                                  Irradiation-induced            ;
radial and axial swelling of the fuel pellets decreases                                                                            ,
the internal volume within the fuel rod.                                                                          In-pile          '
growth of the fuel rod cladding contributes to the internal volume. Axial and radial elastic deformation calculations for the cladding are based on the differential pressure the cladding is exposed to, resulting        in        internal                      volume                                  changes.      Thermal relocation, as well as differential thermal espansion                                                                              ,
of the fuel rod materials also affect the internal volume of the fuel rods.
: 2.            The maximum expected fission gas release in the peak power rod is calculated using the FATES computer code.
Rod power history input to the code is consistent with the design limit peak linear heat rate set by LOCA 4.2-20
 
CESSAR Hmneamu considerations, and therefore the gas release used to                                    ,
size the plenum represents an upper limit.                                Because of    !
time-vary:.ng gap conductance,                            fuel temperature and          :
depletion, and expected fuel management, the release rate varies as a function of burnup.
4.3.1.3.6                      Fuel Rod Performance Steady state fuel temperatures are determined by the FATES                                                        :
computer program.                    The calculations procedure considers the                                    i effect of linear heat rate, fuel relocation, fuel swelling,                                                      l densification, thermal expansion, fission gas release, and clad                                                  I deformation. The model for predicting fuel thermal performance,                                                  j including the specific effects of fuel densification on increased LHGR and stored energy, is discussed in References 15-17.
Significant parameters such as cold pellet and clad diameters,                                                    I gas pressure and composition, burnup and void volumes are                                                      -
I calculated and used as                                    initial  conditions                for  subsequent    l calculations of stored energy during the SIS analysis.                                                    The    )
coupling mechanism between FATES calculations and the SIS                                                        j
  . analysis is described in detail in Reference 26.
Discussions of uncertainties associated with the model, and of
,    comparative analytical and experimental results, are included in
!    Reference 14.
4.2.1.3                      Rurpable Poison Rod i
Two alternative burnable poison rod designs, one using Al 0 -B C poison pellets and the other using Gd,0 -U% poison pelleks,3 ake provided for the System 80 fuel desip.3 BRh types of burnable poison rods have been used previously in C-E-designed reactors and have been approved by the NRC.                                  From the standpoint of fuel              B i
assembly design, the two alternative burnable poison rod designs l    are identical in the cladding material specifications and dimensional properties, and in the mechanical positioning within the fuel assembly.                    The two burnable poison rod designs are described in Section 4.2.2.3. The cladding and burnable poison pellet design properties are described below.
4.2.1.3.1                      Burnable Poison Rod Cladding Design Limits l
l    The burnable poison rod design, similar to the fuel rod design, l
accounts for external pressure, differential expansion of pellets and clad, pellet swelling, clad creep, helium gas release, initial                  internal    helium                pressure,      thermal            stress,  and flow-induced vibrations.                                Except where specifically noted, the design bases presented in this section are consistent with those used for the fuel rod design. The structural criteria for the Amendment B 4.2-21                            March 31, 1988
          .._. _ ._ _ -._. _ -                ~ _ _ _ _ _ . .                _ _ _ _ . _ _ _ _ . _
 
CESSARti & n..
_                l i
1 normal, upset and enorgency loading combinations identified in Sections 4.2.1.1 and 4.2.1.2 are highlighted as follows:
                                                                                                    ]
A. During normal operating and upset conditions, the maximum primary tensile stress in the Zircaloy clad shall not exceed two-thirds of the minimum unirradiated yield strength of the                                  i material at the applicable temperature.      The corresponding                              i limit under emergency conditions is the material yield                                        ,
I strength.
l B. Net unrecoverable circumferential clad strain shall not                                        j exceed 14 as predicted by computations considering clad                                      i creep and poison pellet swelling effects.                                                    i C. The clad will be initially pressurized with helium to an amount sufficient to prevent gross clad deformation under the combined effects of external pressure and long-term                                      ;
creep. The clad design will not rely on the support of pellets or the holddown spring to prevent gross deformation.
4.2.1.3.2        Burnable Poison Rod Cladding Properties.                                        '
Cladding tubes for burnable poison rods are purchased under the                                  ~
specification for fuel rod cladding tubes.          Therefore, the mechanical, metallurgical, chemical, and dimensional properties
* of the cladding are as discussed in Section 4.2.1.2.2.
l 4.2.1.3.3        A1 0 ~I C Burnable Poison Pellet Properties 23  4 The Al 0 -B C burnable poison pellets used        in C-E                designed reactor $3conkist of a relatively small volume fraction of fine B C particles dispersed in a continuous Al O    matrix.              The boron 18ading is varied by adjusting the B C condihnkration in the range from 0.7 to 4.0 wt% (1.0 to 6.0 volk). The bulk density of the A1 0 -B C pellets is specified to be greater than 93% of the ca$c31aked theoretical density. Typical pellets have a bulk density of about 95% of theoretical.      Many properties of the two-phase A1 0 -B C mixture, such as thermal expansion, thermal conductivity,2 3and specific heat are very similar to the properties of the Al o                constituent.        In            contrast, properties such as swelfiflg, major helium release, melting point, and
, corrosion are dependent on the presence of BC            The operating l
centerline temperature of burnable poison pelke.ts        is less than F
1150*F, with a maximum curface temperature of 1090'F.
4.2.1.3.3.1        Thermal-Physical Properties L A. Thermal Expansion The mean thermal expansion coefficients of A1.,0 3              (Reference
: 27) and B4 C (Reference 28) from 0 to 1850'F arn 4.9 and 2.5 Amendment F 4.2-22                December 15, 1989
 
CESSAR til#,c.n.
in./in.
* F x 10 ~0    , respectively.                            The thermal expansion of the Al o -B C two-phase mixture can be considered to be                                                                  j essentdlky h e same as the value for the continuous Al o                                                                  i matrix since the dispersed B,C phase has a lower expanslod
* coefficient and occupies only 5% of the available volume.                                                                ,
The low temperature                  (80 to 250'F) thermal expansion coefficient of Al o irradiated at 480, 900, and 1300'F does not change as a dedult of irradiation (Reference 29) . The expansion of a similar material, beryllium oxide, up to 1900'F, has also been reported to be relatively unchanged by                                                              .
irradiation (Reference 30). It is, therefore, appropriate to                                                              ;
use the values of thermal expansion measured for A1 23                                                0 (Reference 27) for the burnable poison pellets:
Temperature Range                                            Linear Expansion
(*r from 70 to)                                                            ft)                                '
400                                                                0.12                                1 600                                                                0.23                            w 800                                                                0.30 1000                                                                0.40 B.            Melting Point                  .
The melting points of Al o                              (3710*F) (Reference 31) and D C (4400'F) (Reference 32) Erd higher than the melting point 8f                                                            -
the Zircaloy-4 cladding.                          No reactions have been reported                                        ,
l                      between the components which would lower the melting point                                                          . :.
i                      of the pellets to any significant extent. As the B 4 C burns up, the lithium atoms formed occupy interstitia1                                                    sites randomly disrupted within the Bc                                  t  lattice,          rather      than forming a lithium-rich phase (ReTerence 33).                                              The solid solution of lithium in B C should not appreciably influence 4
the melting point of the Al,0,-B4 C pellets, as only a small                                                              ,
quantity of lithium comp 6utids                                    (0.5 wt%) forms during irradiation.        It is concluded that the melting point of Al 0 -B,C will remain considerably above the maximum ll50'F opdr$ ting temperature.
C.            Thermal Conductivity The thermal conductivity of Al o -B C was calculated from the        measured values              for                    Af d 4 and BC using the Maxwell-Eucken relationship (Refdrdnce 34) f8r a continuous matrix phase (Al o            with spherical dispersed phase (B C particles.          Becadsd)of                                      Al        content of th8se) l L                      mixtures and the similarity              the highther$oal    in                conductivity, the resultant values for Al,0 -B              3 C were essentially the same as the values for Al o              The skarured, unirradiated values of thermal conductiv$th. at 750'F are 0.06 cal /s-cm 'K for B4 C l                      and 0.05 cal /s-cm 'K for Al23                      0.
4.2-23
 
i i
l CESSARIMem.
l The                        thermal                    conductivity      of  Al o                after                irradiation decreases rapidly as a function of bdr$up to values of about                                                                                    ;
one-third                                        the unirradiated      values        (Reference                        29}. The          !
irradiated values of Al 0 -B C calculated from the above                                                                                        i relationships are given $elow4 as a function of temperature                                                                                      1 (References 29 and 35).
Temperature                                                        Thermal Conductivity
('F)                                                          fcal/s-am 'K) 400                                                                      0.015 600                                                                      0.013 800                                                                      0.010 1000                                                                          0.008                    !
l    D.                    Specific Heat                                                                                                                                    )
The specific heat of the Al 0 -B C mixture can be taken to                                                                                      l be                      essentially                        the    same 2 $s 4 pure- Al,o                                since the              j concentration of B C is low (6.0 v/o maximum).3 In addition, the effect of irra81ation on specific heat is expected to be                                                                                    l small based on experimental evidence from similar materials which do not sustain transmutations as a function of neutron                                                                                    l exposure.
The                      values                    for Al O      measured    on        unirradiated                        samples (References 35 and 3l)3are given below:
Temperature
__  l'F)                                    cal /mn~ ' F                                      )
l 250                                              0.12                                    '
450                                              0.13                                    1 800                                              0.14                                    l p                                                                                1000 and above                                    0.15                                    )
4.2.1.3.3.2                                                              Irradiation Properties A.                    Swelling A1,0 -B,C consists of BC particles dispersed in a                                                                                                j contlnu5 usa 10                                          matrix, whfch occupies more than 95% of the                                            '
poison pellet. 3 The swelling of Al,0 B C depends primarily 2
upon the neutron flueng on the conti,huoits Al 0 3 matrix and, secondarily, on the B                                              burnup of the dispeded B C phase.
Recent measurements performed on material cong inin,g about 2                                                                                    1 wt% BC                      4                irradicr.ed p a C-E PWR to 100% B                                          burnup at a fluence of 2.4 x 10                                              nyt (E>0.8 MeV) revealed a diametral swelling of about 1%.                                                Pellets similar to the burnable                                            j poison used in C-E reactgs with up to 3 vtt BC also sustained                                          about    100%    B      burnup.                Experimentai data 4.2-24
 
I CESSAR !!M,eu...                                                                                                                                .
I (Reference 0.7%    at a fluence      37) on Al or,0,4            reveal x 10 2 p nyt diametral swelling of about T.                          (E>0.8 MeV).                  Diametral swelling of A10                3 3  I"C#**              * *h  linearly        with      fluence        to 1.8%
after an exposure of 6 x 10                                    nyt E>0.8 MeV).
These data show that A1 0 -B C swells somewhy more than Mek)3. up to~a burnup of h0$ B(about 2 x 10 Al 0                                                                                                    nyt, E>0.8 The C-E Al 0 -B C sw                                    ling rate design value for fluence values lese hah this is, therefore, greater than the swelling rate of Al O , while beyond this threshold the swelling rate for Al 23                        0          C is considered equal to that of                                                  l 4
Al O .
23                                                                                                                                        j These data and considerations result in best-eAtimate diametral swelling values at end-of-life (7 x 10                                                                            nyt, E>0.8 MeV)                  of  about            2%    for    Al  0 23      and      from          2  to      3%          for 0 -B 4 C..
A1 23 w
B. Helium Release                                                                                                                            ,
Experimental measugements reveal that less than 5% of the                                                                                i helium formed during irradiation will be released (Reference.
38). These measurements were performed on Al 0 -B,C pellets irradiated                  at  temperatures                  to    500'F        and,2 dub 5equently, annealed at 1000'T for 5 days.                                          The helium release in a a
burnable poison rod which operated for one cycle in a C-E                                                                            '
PWR was calculated from internal pressure measurements to be                                                                              ,
less than 5%.
4.3.1.3.3.3                        Chemical Properties                                                                                        i A. Al 23 0 -B C4 Coolant Reaction Should    irradiated                    BC particles                    be        exposed            to        reactor coolant, the primary 8errosion products that would be formed                                                                              i-are boric acid (which is soluble in water), hydrogen, free                                                                                '
carbon    and a small amount of lithium compounds.                                                                          The presence of these products in the reactor coolant would not be detrimental to the operation of the plant.
Observations of Al 0 -B3 C poison shims nave revealed that                                                                                ,
long term exposure                    2 od tnis material to reactor coolant can result in gradual leaking out of boron and eventual eroding away of the Al o                    matrix.                However, the rate of reaction is such that any2 desultant changes in reactivity are very gradual.
4.2-25
 
i t
CESSAR IB#icm a f
0 B. Chemical Compatibility C pellets and the Chemical compatibility burnable            poison rod                      between        Al 03-B'long-term normal claddingthedudnD operation has been demonstrated by examination of a burnable                                                                      '
poison rod from the Maine Yankee reactor. The rod had beg exposed to an aximuy averaged fluence in cxcess of 2 x 10 nyt (E>0. 821 NeV) .                              No evidence of a chemical reaction was observed on the cladding ID.
Short-term chemical compatibility during upset and emergency conditions            is demonstrated by the fact that conditions                                                                !
favorable to a chemical reaction between Zircaloy-4 and Al o          are not present at temperatures below 1300*F                                                                        ,
(R$fdrence 39). This temperature is higher than that which will occur at burnable poison pellet surfaces during e
* Condition II and III occurrences (Section 4.2.1.1).                                                        The
:            reaction between Zircaloy-4 and A1 0                                          described by Idaho                                  t 3
Nuclear (Reference 40) was observed Eo                                      occur rapidly.only at                                ,
temperatures in excess of                                    2500'F, well above the peak Zircaloy-4 temperatures                                  in    the higher-energy                fuel rods described in Chapter 15.                                                                                                          '
4.2.1.3.4                  Gog3o -Do                    Burnable Poison Pellet Properties 3
This      section          references                      evaluations      of    gadolinia-urania                                  ,
properties and of thermal conductivity and melting temperature                                                                          l correlations appropriate for gadolinia-urania compositions of                                                                          l B
interest in PWR applications of Gd 0 -Uo burnable absorbers.
23    2 The material properties that influence the thermal performance of gadolinia trania fuel have been reviewed to ascertain how UO properties are influenced by the addition of gadolinia.                                                        Thes$                      l include the thermal conductivity, solidus temperature, specific heat, and the coefficient of thermal expansion. The effects of gadolinia addition on these properties are discussed in detail in Reference 41.
4.2.1.4              control Element Assemblies fcEAs)
Except where specifically noted, the design bases presented in this section are consistent with those used for previous d6aigns.                                                                        ,
The mechanical design of the control element assemblies is based on compliance with the following functional requirements and criteria:                                                                                                                                l l
A. CEAs will provide for or initiate short-term reactivity control under all normal and adverse conditions experienced during reactor startup, operation, shutdown, and accidents.
1 Amendment B                                      1 4.2-26                      March 31, 1988                                    )
 
C E S S A R 83 & .r,..
B. Mechanical clearances of the CEA within the fuel and reactor                            i internals are such that the requirements for CEA positioning                              i and reactor trip are attained under          the    most  adverse accumulation of tolerances.
C. Structural    material    characteristics    are    such    that radiation-induced changes to the CEA materials will not                                  ;
impair the functions of the reactivity control system.                                    l i
4.2.1.4.1        Thermal-Physical Properties of Absorber Material The absorber materials used for full-strength control rods are boron carbide pellets _ (3 C) and silver-indium-cadmium bars                F                1 (Ag-In-Cd). Inconel Alloy k25 is used as the absorber material                              l for the part-strength control rods.          Refer to Figures 4.2-3,                            i 4.2-4, and 4.2-5 for the specific application and orientation of the absorber materials.      The significant thermal and pitysical properties used in mechanical analysis of the absorber materials                        ;      )
are listed below:                                                                      .w i
A. Boron Carbide (B 4C)                                          ,                      .
Configuration              Right cylinder Outside Diameter,          (a) 0.737'10.001 inches                      (b) 0.664 10.001 (reduced diam.)
Pellet Length,              (a) 2.00                                              ~
l inches (nominal)            (b) 1.79 (reduced diam.)
l                    End Chamfer                0.007 to 0.020 Radius, inchas Density, Ib/in 3            0.066 Weight % Boron,            77.5 minimum l
4 Open Porosity in          27                                                          ,
Pellet
!                    Ultimate Tensile            N/A Strength, Ib/in.2
[
Yield gtrangth, Ib/in.                      N/A Elongation, %              N/A Young's Modulus, psi        N/A Amendment F 4.2-27                  December 15, 1989
 
CESSAR timnem..                                                                                                                                    1 i
Thermal conductivity                                          Irradiated          Unirradiated                                        )
(Btu /hr-ft 'F):
800'F                                            2.0                6.8 1000'F                                            1.9                5.8 Melting Point, 'F                                            4,440-4 Linear Thermal                                                                                                                      :
Expansion                                                    0.234 0 1000*F                                                          ]
1 B.          Type 347 Stainless Steel                                                                                                              j (Felt Metal)                                                                                                                        ;
Configuration                                                cylindrical sleeves                                                      !
formed from sheets                                                      ;
Thickness, inches                                            0.032 1 0.002 0
Length of Sheet,                                              12.34-inches (nominal)                                                                                                                      '
Density, lb/in.3                                              0.059
    .                                                                                                                                                  i Ultimate Tensile                                              N/A                                                                      I Strength, lb/in.2                                                                                                                      ,
Elongation, %                                                N/A Yeung'g Modulus,                                                                                                                      i lb/in.                                                        N/A                                                                      !
l Thermal Conductivity                                                                                                                  ;
(Btu /hr-ft 'F)                                                                                                                      1 500*F                                            0.30 1000'F                                            0.34                                                                    ';
C.          Silver-Indium-Cadmium (Ag-In-cd)
Configuration                                                cylindrical bars with central hole                                                                f  ,
Outside Diameter, in.                                        0.734 1 0.003 Inside Diameter, inches                                      0.25 Length of Bar, inches                                        2 nominal l
Amendment F 4.2-28                    December 15, 1989
 
CESSAR IMric r.
i Density, lb/in.3                                                      0.367                                                                              I Ultimate Tensile                                                      N/A Strength, lb/in.2                                                                                                                                        t Yield Strength, lb/in.2                                                373 Elongation,,4                                                          N/A Young's Modulus, lb/in.2                                              N/A                                                              ,              f Thermal Conductivity                                                  Irradiated                              Unirradiated (Btu /hr-ft *F)I                                                                                                                                        j 572*F                                                      34.                                      44.
752*F                                                      36.                                      47.
t Melting Point, 'F                                                      1,470 F
                                                                                                                        -6 Linear Thermal Expansion                                              12.5 x 10                                                                        ,
(in./in. *F)
D.      Inconel 625 (Ni-Cr-Fu Alloy)                                                                                                                            ,
configuration                                                          Solid cylindrical bars (as absorber)
Diameter, inches                                                      0.737 1 0.001 Length of bar, inches                                                  2 nominal Density, lb/in.3                                                      0.305                                                                            -
Ultimate Tensile                                                      120-150 Strength, lb/in.3 Specified Minimum Yield                                                65 Strength 9 650'F, kai Elongation in 2 inches, 4                                              30                                                                                .
l Young's Modulus, lb/in.2 6
70'F                                                        29.7 x 10 650*F                                                            27. 0 :: 10 Amendment F 4.2-29                                                            December 15, 1989 L -  . _. _      . _ . -    . _ . _ . . . . _ . . . . _ . . . _ _ _ _ . . . . . _ _ . - . _ - . _ . _ . - . _ _ _ . .      _ _ _ _ _ . . . _ __      _ . . _ . -
 
C E S S A R U W ico..
l Thermal Conductivity (Stu/hraft 'F):
70'F                    5.7 600'F                    8.2 Linear Thernal Expansion                    -6 7.4 x 10 (in./in. 'F)                        (70 to 600*F)
F 4.3.1.4.2                Compatibility of Absorber and Cladding Materials' The cladding material used for the control elements is Inconal Alloy 625. The selection of this material for use as cladding is based on consideration of stren@, creep resistance, corrosion resistance, and dimensional stability under irradiation, and upon                    .j the acceptable performance of this material for this application in other c=E reactors currently in operation.
A.        84 C/Inconal 625 Compatibility ctudiaa have been conducted cy HrDL (Reference 42) on the compatibility of Type 316 stainless steel with EC under 4
* irradiation for thousands of hours at tomoeraturer between 1300 and 1600*F. Carbide formation to a depth of about 0.004 inches in the steel was measured after 4400 hours at 1300'F.
Einilar compound formation depths were observed after ex-reactor bench testing.          After testing at 1000'F, only 0.001 in./yr of penetration was measured. Since Inconel 625 is more resistant to carbide formation            than Type 316 stainless steel, and the expected pellet / clad interfacial temperature in the standard design is below 800'7,                it is concluded that B C is compatible with Inconel 625.
4 F
4 2 1.4.3                Cladding stress-strain Limits The stress limits for the Inconel Alloy 625 cladding are as follows:
Amendment F 4.2-30              December 15, 1989 I
 
1 l
CESSAR tilhi.                                                                                                                      !
1 Design for Non-operation, Normal Operation, and Upset Conditions:
P, 5 S, l
P, + Pb s F,S, The not unrecoverable circumferential strain shall not exceed 1%                                                                  i on the cladding diameter,                    considering the effects of pellet                                                    l l
swelling and cladding creep.
Design for Energenev Conditions:
i P,s 1.5 S, P, + Pb s 1.5 F,S,                                                                                            1 Design for Faulted Conditions:                                                                                              ,
r P,5 S',                                                                                                      ,
P, + Pb $ F,S, whereSgisthesmallerof2.4S,or0.7S'                            u For definition                  of P    P  S    S'    S  a                F,                  see Section For the  IEc,ond,  62E,CEA        E,laddng,ndthe,value                                      of S*
4.2.1.1.1.                                                                                                              is two-thirds                  of  the  minimum    specified    yield                            strength                        ;
temperature.                                                                                                                ,
For Inyonel 625, the specified minimum yield strength is 65,000 l
lb/in.                  at 650'F.
i            F          = Mp/My where Mp is the bending moment required to produce a                                                          -
f611y plastic section and My is the bending moment which first produces yielding at the extreme fibers of the cross-section.                                                                    ,.
The capability of cross-sections loaded in bending to sustain accents considerably in excess of that required to yield the outermost fiber is discussed in Reference 1.                                                    For the CEA cladding dimensions, F, = 1.33.
The values of uniform and total elongation of Inconel Alloy 625 cladding are estimated to be as follows:
i-Fluence.(E >1 MeV), nyt l'                                                          1 x 10''          3 x 10''
Uniform elongation, %          3                            1 Total elongation, 4            6                            3 4.2-31 l.
                                                                  . _ ~      _ . _ _ _ _ . _ . _ . _ _ _ _ _ _ _ _ _ . - . - _ -
 
i l
CESSAR IIMcui.
4.2.1.4.4                    Irradiation Behavior of Absorber Materials                                                                                  j A.          Boron Carbide                                                                                                                                  i i
: 1. Swelling                                                                                                                                  ,
The      linear                swelling      of        B                  increases            with          burnup according to the relationship,Cs 0
                                %AL = (0.1) B                      Burnup, at %
This        relationship was                        obtained from experimental irradiations                        on    high      density                                  (k90% TD)            wafers (Reference 43) and pellets with densities ranging between 71 and 98% TD (References 42 and 45).
Dimensional changes were measured as a function of burnup, after irradiating at temperatures expected in the design.
: 2. Thermal conductivity                                                                                                                    :
                                                                                                                                              ~
The thermal conductivity of.unirradiated 73% dense B C decreases linearly with' temperatures from 300 ko 1600'F, according to the relationship:                                                                                                ,
1 cal /s-en 'E 2.17 (6.87 + 0.017 T)
This relationship was obtained from measurements                                                                                        ,
performed on pellets ranging                                                        from 70 to 98% TD (Reference 45)..
The relationship between the thermal conductivity of                                                                                    '
i                    irradiated 73% TD BC pellets and temperature given below was derived frba measured values (Reference 45) onhiggrdensitypelletsirradiatedtofluencesoutto 3 x 10          nyt (E>l MeV):
L                                  "
1 cal /s-em *K 2.17 (38 + 0.025 T) where T - temperature,                        'K.
Thermal conductivity measurements of 17 BC specimens                                                                                    ,
with densities ranging from 83 to 98% TD, ibradiated at temperatures from 930 to 1600*F showed that thermal conductivity decreased significantly after irradiation.
The rate of decrease is high at the lower irradiation temperatures, but saturates rapidly with exposure.
4.2-32 l
 
CESSARRELn.                                                                                            l I
: 3. Helium Release                                                                                  j 10 Helium is formed in B C                  4 as B      burnup progresses.                  The fraction    of  helium                    released from the pellets is
            .important for determining rod internal gas pressure.                                            l The relationship between helium release and irradiation temperature      given                    below was developed                          at  ORNL (Reference 46) to fit experimental data obtained from thermal reactor irradiations:
                                                                    "Q                                        \
                  % He release = e(C-1.85D),RT                                                              J where C =  Constant, 6.69 for pellets D  =  Fractional density, 0.73 for C-E pellets                                            -
Q =  Activation energy content, 3600 cal / mole R =  Gas constant, 1.98 cal / mole *K T =  Pellet temperature, 'K This expression becomes                                                                        j
                                                            .-1820
                  % He release = 208 e T                          +5                                      .
When the above parameters are substituted.                                              In this form, design values for helium release as a function of temperature are generated.                            The 5% helium release allowance (the last term in the expression) was added-                                          >
to ensure that design values lie above all reported helium    release data.                      Calculated values of helium release obtained from the recommended design expression lie above all experimental data points (Reference 42, 47, 48) obtained on B 4 C pellet specimens irradiated in thermal reactors.
: 4. Pellet Porosity Experimental evidence is available (Reference 49) which shows that for pellet densities below 90%, essentially
* all    porosity                is          open      at    beginning-of-life, i            Irradiation-induced                      swelling      does not change the characteristics of the porosity, but only changes the
!          bulk volume of the specimens. Therefore, the amcunt of porosity available at and-of-life is the same as that present at beginning-of-life.
4.2-33
 
t i
CESSAR Mne.                                    .
l                                                                                                                                                    1 B. Silver-Indium-Cadmium
: 1. Swelling Measurements performed                                                                    2  gn Ag-In-cd rods irradiated at        i fluences up to 6.2 x 10                                                                        nyt (E >0.6 eV) were employed to develop the following expression to predict the                                                                                  ,
volumetric swelling for silver-indium-cadmium alloy:
                          %AV = 21                        U                                                                                          l l                                              10 where d = fluence, nyt (E >0.6 eV).
Linear swelling is                                                                      approximately      one-third  of    the volumetric swelling.
: 2. Thermal conductivity                                                                                                                ?
The increase in cadmium content from 5 to perhaps 10 wtt, and the formation of 2 to 3 wt% tin as a result of                                                                              '
l                long-term exposures,                                                                      is    expected to decrease the thermal conductivity from the expected to decrease the thermal conductivity from the accepted (Reference 50)                                                                              F unirradiated values.                                                                      Published data for unirradiated Ag-cd binary alloys shows that thermal conductivity was decreased                    by                    about                              20%    by  increasing  the  cadmium      ,
content from 5.0 to 10.0 wt% (Reference 50).                                                                              Sinca irradiated Ag-In-Cd is expected to perform in much the same      fashion, similarly, the unirradiate values of thermal conductivity are decreased by 25% to account for irradiation.
: 3. Linear Thermal Expansion                                                                                                            '
The coefficient of lianar thermal                                                                              expy sion for unirradiated Ag-In-Cd material is 12.5 x 10                                                                          in/in. *F over the temperature range of 70 to 930'F (Reference                                                                                  ,
51). Published data on unirradiated (Reference 50)
: l.              Ag-cd binary alloys reveal that a cadmium incre.ase of 1                5% will result in about a                                                                            5% increase in thermal expansion coefficient.                                                                      The small changes in indium and tin content do not influence the thermal expansion coefficient appreci,, ply.                                                                      For simplicity, irradiated I                values of 13.1 x 10                                                                    in./in.    'F are used in all design calculations.
Amendment F 4.2-34                              December 15, 1989
                            . . . - - - - . . . . . - _ . - . - . - - . . _ . . . . . ~ . - - . . - , .
 
CESSAR E h .,
i i
: 4.        Mel*cing Point i
The melting point of unirradiated Ag-In-Cd has been measured as        1470*F 1                          30'F    (Reference 50).                The formation of 3 wt% tin due to the transmutation of indium and the increase in cadmium content to about 10 wtt due to the transmutation of silver may result in a                                                            !
small decrease in the melting point.                                                                              l C. Inconel 625
: 1.        Swelling Available information indicates that Inconel 625 is                                                                l highly resistant to radiation swgling. Exposure of Inconal 625 to a fluence of 3 x 10                                nyt (E>0.1 MeV)'at I
a temperature of 400'C                                (725'F) showed no visible cavities in metallographic examinations (Reference 52)                                                            ;
so that swelling, if any, would be very minor. Direct measurements mag after exposure of Inconel 625 to a
* fluence 5 x 10                nyt (E>0.1 MeV) at LMFBR conditions                                    .
showed no evidence y swelling (Reference 53). Further                                                -
      .          exposure to 6 x 10                            nyt (E>0.1 MeV) at 500*C (932'F)                                    ;
showed essentially no swelling as measured by immersion density, but did show small cavities. Thus, Inconel                                                                ,
62$2 is n t expected to swell beyond fluences of 3 x                                                          ,
10      nyt (E>l MeV).                                                                                          ,
: 2.        Ductility                                                                                                    *
                .The      ductility          of                  Inconel    625      decreases              after irradiation.      Extrapolation of lower fluence data on Inconal 625 and 500 indicates that the values of unyorm and total elongation of Inconal 625 after 1 x 10      nyt (E>l MeV) are 3 and 6%, respectively.
: 3.        Strength The value of yield strength for Inconel 625 increases                                                              ,
after irradiation in the manner typical of metals.
However, no credit is taken for increases in yield strength in the design analyses above the value initially specified.
4.2.1.5                surveillance Procram 4.2.1.5.1                Requirements for Surveillance and Testing of Irradiated Fuel Rods High burnup performance experience, as described in                                                        Section 4.2.3,        has provided evidence that the fuel will                                                    perform F Amendment F 4.2-35                              December 15, 1989
 
CESSAR Enfiem.                                                                                                                    i i
satisfactorily                                under        the  design    conditions. Two  irradiated programs were developed for fuel performance surveillance in                                                                    '
Arkansas Nuclear One-Unit 2 (ANO-2). The fuel rods in these 16x16 fuel assemblies are similar to those in the System 80 design.
The first fuel performance program in ANO-2 has been completed.
This program followed six standard assemblies through three I
irradiation cycles.                                      Each assembly contained pre-characterized fuel rods which were examined during refueling shutdowns.                                                  The results of the program demonstrated that the fuel assemblies performed reliably through overaged burnups of 37.2 GWd/MTU.                                                                      ,
Zircaloy oxide thicknesses, fuel rod growth and bowing, and i
assembly dimensional stability were consistent with initial predictions (Reference 54).
The second program, which is nearing completion at ANO-2 has irradiated two fuel assemblies containing both standard and                                                                    I advanced design fuel rods to extended burnups.                                          Both assemblies were extensively pre-characterized. One assembly was irradiated for three reactor cycles and reached an assiembly-averaged burnup of 33 GWd/MTU.                                  A second assembly was exposed to 5 cycles and reached an assembly-averaged burnup of 52 GWd/MTU (Reference 55).
Both assemblies were examined after each reactor cycle.                                                Visual examinations, oxide thickness measurements, and other dimensional measurements results in the conclusion that the performance of                                                                    ,
the          fuel                has                been    satisfactory.      Destructive    hot      cell examinations are scheduled to complete the characterization of fuel behavior.
A surveillance program to follow the fuel performance of the System 80 design is being carried out in Palo Verde-1.                                                    The program includes poolside examinations after each of the first three            operational                          cycles. The    examinations  include        visual inspections for overall performance, dimensional measurements to                                                                  '
characterize growth behavior, and cladding oxide measurements to                                                                  ,
track corrosion behavior of the fuel rod cladding.                                          Results after the first and second cycles indicate that the fuel is behaving as expected with no indication that would alter the planned fuel management scheme for the System 80 fuel.                                            The final poolside examination of these standard assemblies is scheduled for 1991 when they will be discharged.
4.
 
==2.2                            DESCRIPTION==
AND DESIGN DRAWING 8 This subsection summarizes the mechanical design characteristics of the fuel system and discusses the design parameters which are of significance to the performance of the reactor. A summary of mechanical design parameters is presented in Table 4.2-1. These data are intended to be descriptive of the design; limiting Amendment F 4.2-36                  December 15, 1989
 
F CESSAR Mncan.
i values of these and other parameters will be discussed in the                                      '
appropriate sections.
4.a.a.1          Puel AssanMy The fuel assembly (Figure 4.2-6) consists of 236 fuel and poison rods, 5 guide tubes, 11 fuel rod spacer grids, upper and lower end fittings, and a holddown device.                        The outer guide tubes, spacer grids, and and fittings ' form the stractural frame of the assembly.
The fuel spacer gride (Figure 4.2-7) maintain the fuel rod array by providing positive lateral restraint to the fuel rod but only frictional restraint to axial fuel rod motion. The grids are                                                  r fabricated from preformed Zircaloy or Inconal strips (the bottom spacer grid material is Inconel) interlocked in an egg crate fashion and welded together.                Each cell of the spacer grid                                    ,
contains - two leaf springs and four arches.                      The leaf springs press the rod against the arches to restrict relative motion                                          '
between the grids and the fuel rods.                          The perimeter strips                          !'
contain features designed to prevent hangup of grids during a refueling operation.                                        ,
The ten Zircaloy-4 spacer grids are fastened to the Zircaloy-4 guide tubes by welding, and each grid is welded to each guide                                                .
tube at eight locations, four on the upper face of the grid and                                              ?
I four on the lower face of the grid, where the spacer strips                                                  .
contact the guide tube surface. The lowest spacer grid (Inconel)                                            l is not welded to the guide tubes due to material differences. It is supported by an Inconel 625 skirt which is welded to the l
spacer grid and to the perimeter of the lower and fitting.
The upper and fitting is an assembly consisting of two cast Type 304 stainless steel plates, five machined posts and four helical Inconel X-750 springs.              The upper and fitting attaches to the guide tubes to serve as an alignment and locating device for each fuel assembly and has features to permit lifting of the fuel assembly. The lower cast plate locates the top ends of the guide tubes and is designed to prevent excessive axial motion of the fuel rods.
The Inconel X-750 springs are of conventional coil design having                                  g a mean diameter of 1.859 in., a wire diameter of 0.319 in., and approximately 16 active coils.              Inconel X-750 was selected for this applicatien because of its previous use for coil springs and good resistance to relaxation during operation.
The upper cast plate of the assembly, called tha holddown plate, together with the helical compression springs, comprise the l            holddown device.        The holddown plate is movable, acts on the                              B underside of the extended tubes of the upper guide structure and l
Amendment B 4.2-37                      March 31, 1988
 
I CESSAR W h mn 4
is loaded by the compression springs.                Since the springs are located at the upper end of the assembly, the spring load                              I combines with the fuel assembly weight to counteract upward                            I
;    hydraulic forces.      The determination of upward hydraulic forces                    ;
includes factors accounting for flow maldistribution,                    fuel          l assembly component tolerances, crud buildup, drag coefficient, and bypass flow. The springs are sized and the spring preload is selected such that a net downward force will be maintained for
    -  all normal    and anticipated transient flow and temperature conditions. The design criteria limit the maximum stress under the most adverse tolerance conditions to below the yield strength
,      of _ the spring material. The maximum stress occurs during cold                    1 l      conditions and decreases as the reactor heats up. The reduction                        I i      in stress is due to a decrease in spring deflection resulting                          l l
from differential thermal expansion between the Zircaloy fuel assemblies and the stainless steel internals.
!      During normal operation, a spring will never be compressed to its solid height. However, if the fuel assembly were loaded in an abnormal manner such that a spring were compressed to its solid                        j l    height, the spring would continue to serve its function when the                        !
j      loading condition returned to normal.                                                  l l
The lower and fitting assembly is a simple stainless                      steel i      casting consisting of a plate with flow holes and a support leg i
      .at each corner (total of four legs) that aligns the lower end of the fuel assembly with the core support structures alignment pins. Each alignment pin is required to position the corners of four lower and fittings.        A center post is threaded into the central portion of the flow plate and welded into position.
The four outer guide tubes have a widened region at the upper end which contains an internal thread. Connection with the upper end fitting is made by passing the externally threaded and of the guide posts through holes in the lower cast flow plate and into the guide tubes.      When assembled, the flow plate is secured between flanges on the guide tubes and on the guide posts.                  The connection with the upper and fitting is locked with a mechanical crimp. Each outer guide tube has, at its lower end, a welded I
Zircaloy-4 fitting. This fitting has a threaded portion which passes through a hole in the fuel assembly lower end fitting and is secured by a Zircaloy-4 nut.        This joint is secured with a stainless steel locking disc tack welded to the lower end fitting in four places.                                                                B The center instrumentation guide tube inserts into a socket and a sleeve in the upper and lower end fittings, respectively, and is thus retained laterally by the relatively small clearance at i    these locations. The upper and fitting socket is created by the
,    center guide tube post which is threaded into the lower cast flow l
l                                                                  Amendment B 4.2-38                      March 31, 1988
 
CESSAR!nMe                                        .
plate and tack welded in four places.                          The lower end fitting  B sleeve is an extension from the center post of the lower end fitting assembly.                        There is no positive axial connection between the center guide tube and the and fittings.
The five guide tubes have the effect of ensuring that bowing or excessive swelling of the adjacent fuel rods cannot result in obstruction of the control element pathway. This is so because:
A.      There is sufficient clearance between the fuel rods and the guide tube surface to allow an adjacent fuel rod to reach rupture strain due to excessive swelling without contacting                    B the guide 'ube c                          surface.
B. The guide tube, having considerably greater diameter and wall thickness (and being at a lower temperature) than the fuel rods, is considerably stiffer than the fael rods and would, therefore, remain straight, rather than be deflected                      g by contact with the surface of an adjacent bowed fuel rod.
Therefore, the bowing or swelling of fuel rods would not result in obstruction of the control element channels such a r. could hinder CEA movement.
The fuel assembly design enables reconstitution, i.e., removal and replacement of fuel and poison rods, of an irradiated fuel assembly. The fuel and poison rod lower end caps are conically shaped to ensure proper insertion within the fuel assembly grid cage structure; the upper end caps                        are designed to enable grappling of the fuel and poison rod for purposes of removal and handling. Threaded joints which mechanically attach the upper end fitting to the control element guide tubes will be properly torqued and locked during service, but may be removed to provide access to the fuel and poison rods.
Loading and movement of the fuel assemblies is conducted in accordance with strictly monitored administrative procedures and, at the completion of fuel loading, an independent check as to the location and orientation of each fuel assembly in the core is required.
The serial number provided on the fuel assembly upper end fitting enables verification of fuel enrichment and orientation of the fuel assembly. The serial number is also provided on the lower end fitting to ensure preservation of fuel assembly identity in the event of upper end fitting removal. Additional markings are provided on the fuel rod upper end caps as a secondary check to distinguish between fuel enrichments and burnable poison rods, if present.
Amendment B 4.2-39            March 31, 1988
 
C E S S A R W #icui.
i 1
l I
During the manufacturing process, .the lower and cap of each rod                  l is  marked to provide a means of identifying the pellet                            i enrichment, pellet lot and fuel stack weight.                In addition, a      i quality control program specification requires that measures be                    l established for the identification and control of materials,                      ;
components, and partially fabricated subassemblies. These means-provida- assurance that only acceptable items are used and also                    ,
provide a method of relating an item or assembly from initial                      !
receipt      through    fabrication,      installation,      repair,      or  !
modification to an applicable drawing, specification or other                      j pertinent technical document.                                                      1 4.2.2.2          Fuel Rod
:i The  fuel      rods consist of        slightly-enriched UO        cylindrical ceramic      pellets,  a      round  wire  Type      302  stdnless      steel compression spring, and an alumina spacer disc located at each
,  and of the fuel column, all encapsulated within a Zircaloy-4 tube seal welded with Zircaloy-4 end caps.                    The fuel rods are
;  internally pressurized with helium during assembly.              Figure 4.2-8 depicts the fuel rod design.
Each fuel rod assembly includes both a serial number and a visual identification mark. The - serial number ensures traceability of the fabrication history of each fuel rod component.                          The identification mark provides a visual check on pellet enrichment batch during fuel assembly fabrication.
The fuel cladding is cold worked and stress relief annealed Zircaloy-4 tubing 0.025 inches thick.              The actual tube forming process consists of a series of cold working and annealing operations, the details of which are selected to provide the combination of properties discussed in Section 4.2.1.2.2.
J
,  The UO      pellets are dished at both ends in order to better
[  accomm8date thermal expansion and fuel ayelling.              The density of i
the UO 2 in the pellets iry        10.38 g/cm , which corresponds to 94.75% of the 10.96 g/cm              theoretical density      (TD)    of UO .
However, because the pellet dishes and chamfers constitute abodt 3% of the volume of the pellet stack,3the average density of the                  1 pellet stack is reduced to 10.06 g/cm .              This number is referred to as the " stack density".                                                        l I
The compression spring located at the top of the fuel pellet column maintains the column in its proper position during                          i
;  handling and shipping. The alumina spacer disc at the lower end of the fuel rod reduces the lower end cap temperature, while the
;  upper spacer disc prevents UO, chips, if present, from entering j  the plenun region.      The      f.lel rod plenum,    which is located above l
l l
4.2-40 l
 
1 m
CESSAR Blubn.                                                                                                                j the pellet column, provides space for axial thermal differential expansion of the fuel columet and accommodates the initial helium loading and evolved < fission gases.                              (see Sections 4.2.1.2.5.1 and 4.2.1.2.5.2.)              The- s;pecific manner in which these factors are taken into account, . ancluding the calculation of temperatures for                                                          ,
the gas contained within'the various types of rod internal vald volume, is discussed-in References 15-17.
            ' 4 ~. 2 . 2 . 3            Burnable Poison Rod Fixed' burnable neutron absorber (poison) rods will be included in selected            fuel -assemblies to                reduce              the  beginning-of-life moderator coefficient.                      They will replace fuel rods at selected' locations.            The two alternative burnable poison- rod designs are                                    B described below.                                                                                                              '
A.      'A123    0 - B 4C Burnable Poison Rod                                                                                ,
The poison rods shown in Figure 4.2-9 will be mechanically similar to fuel rods, but will contain a column of burnable                                              "
['
poison pellets instead of fuel pellets. The poison material will be alumina with uniformly-dispersed boron. carbide particles.              The balance of the column will consist of Zircaloy-4 spacers,'with the total column length the same as the column length in fuel rods.                                  The burnable poison rod                    ',
plenum spring is designed to produce a smaller preload on the pellet column than that in a fuel rod because of the lighter material in the poison pellets.
B.        Gd23    0 -
0
: 2. Burnable Poison Rod                                                            B The poison rods.shown in Figure 4.2-10 will be mechanically L                      similar to fuel rods, but will consist of Gd.30 admhed in natural UO 2 in the central rod portion (axialli)3 and natural UO, at the top-and bottom. The total column lcagth is the same as the column length in fuel rods.
Each burnable poison rod assembly includes a serial number and visual identification mark. The serial number is used to record fabrication information for each component in the rod assembly.
The' identification mark is unique to poison rods and provides a
            . visual check on the pellet poison content during fuel bundle                                                  B fabrication.
Amendment B 4.2-41                              March 31, 1988
 
CESSARI!ninen.
4.2.2.4        control Element Assembly Descrintion and Desian L
The  control neutron      element absorber        assemblies elements  arrangedconsist to engage    of either  four or twelve the-peripheral  guide lF tubes of fuel assemblies.        The neutron absorber elements are                        l connected by a spider structure which couples to the control-element drive mechanism (CEDM) drive shaft extension.                      The neutron absorber elements of a. four-element CEA engage the four corner guide tubes in a single fuel assembly. The four-element                    B CEAs are used for control of power distribution and core reactivity in the power operating range. The twelve-element CEAs engage the four corner guide tubes in one fuel assembly and the two nearest corner guide tubes in adjacent fuel assemblies. The twelve-element CEAs make up the balance of the control groups and provide h bank of strong shutdown rods.                The control element
    -assemblies are shown in Figures 4.2-3 through 4.2-5. The patterna of CEAs (total of 93) is shown in Figure 4.2-11.. Note that up to lB eight additional CEAs may be installed if desired for additional flexibility or future use.        Twenty-five of the 93 CEAs are p part-strength CEAs (PSCEAs).
Part-strength'CEAs are differentiated from full-strength CEAs by using alphanumeric serialization instead of the numerical system used on the full-strength CEAs.                                                  B All' control elements are sealed by welds which join the tube to an Inconal 625 nose cap at the bottom,                and an Inconel 625 connector at the top which makes up part of the end fitting.                The end fittings, in turn, are threaded and crimped in place by a locking nut to the spider structure which provides rigid lateral and axial support for the control elements.              The spider hub bore is specially. machined to provide a point of attachment for the CEA extension shaft.
The control elements of a twelve-element full-strength CEA consist of an Inconal 625 tube loaded with a stack of cylindrical absorber pellets. The absorber material consists of 73% TD boron carbide (B c) pellets, with the exception of the lower portion of the elemenks, which contain reduced diameter B C pellets wrapped                            l in a sleeve of Type 347 stainless steel (falt aktal).                                      I 1
The  design objective realized by the use of felt metal and                              I reduced B C pellets in the element tip zones is that as the B C                            i pellets Avell due to        irradiation,      the      felt metal    slee6e              '
compresses as a result of the applied loading. This compression limits the amount of induced strain in the cladding. Therefore, buffering of the CEA following scram, which occurs when the element tips enter a reduced diameter portion of the fuel assembly guide tubes, is not affected with long term exposure of the CEA to reactor operating conditions.
Amendment F 4.2-42                        December 15, 1989
 
y            -.              -        -        -  -    -.      .    - --
CESSAR EliMeami.
B During normal powered operation, all of tne twelve-element CEAs are expected to be in the fully withdrawn position. Thus, the local B-10 burnup progresses at a lower rate, and CEA life is prolonged. Above the poison column is a plenum which provides expansion volume for helium released from the B C. The plenum volume contains a Type 302 stainless steel holddo                  spring, which
:  restrains the absorber material against longitudinal shifting with respect to the clad while allowing for                        differential expansion between the absorber and the clad. The spring develops L  a load sufficient to maintain the position of the absorber material during shipping and handling.
L  The control elements of a four-element full-strength CEA consist l  of an Inconel    625 tube loaded with a stack of cylindrical
!  Ag-In-Cd absorber bars.              This CEA design is used for the j  regulating banks.        Two design objectives are realized by use of Ag-In-Cd absorber over the full active length:                                      ,
l  A. CEA Cladding Dimensional Stability                                                  C-l_                                                                                        F  t
(.      Because of its high ductility and low strength, the Ag-In-Cd will not deform the CEA cladding.                  Buffering of the CEA following scram, which occurs when the corner element tips enter a reduced diameter portion of the fuel assembly guide tubes, is not degraded with long-term exposure of the CEA to reactor operating conditions.
B. Adequate CEA Worth                                                                  _
Although some reduction in CEA worth arises because of the substitution of B C with Ag-In-Cd, the effect is small and is accounted for,4 The control elements of the four-element PSCEA consist of an Inconel 625 tube loaded with Inconel 625 bars over the full active length. The PSCEA, which have lower worth in comparison to the full-length CEAs, are provided for reactivity and axial power shape control during power operations. Because of the use of Inconel 625 absorber, the cladding dimensional stability is not degraded with long-term exposure of the PSCEA to reactor operating conditions.
Each full-strength or part-strength CEA is positioned by                          a magnetic jack control element drive mechanism (CEDM) mounted on the reactor vessel closure head. The extension shaft joins with the CEA spider and connects the CEA to the CEDM.                      Full- and        B part-strength CEAs may be connected to any extension                      shaft depending on control requirements.              Mechanical reactivity control is achieved by positioning groups of CEAs by the CEDMs.
Amendment F 4.2-43                  December 15, 1989
                                                                                                ~
 
                                                      --    -  -      -                -      -        a    . . . ~ . -~    +,-a - a
:CESSAR REnem.                                                                                                              !
1 i
I
              'In the outlet plenum region, all CEAs/PSCEAs are enclosed in CEA                                                      B l
shrouds which provide guidance and protect the CEA/PSCEA and extension shaft from coolant cross flow.                                        Within the core, each                    l element travels in a Zircaloy guide tube.                                          The guide tubes are                  i part of the: fuel assembly structure and ensure proper orientation of=the control elements with respect to the fuel rods.
l When the extension shaft is released by the CEDM, the combined weight of the shaft and CEA causes the CEA to insert into the fuel assembly.
The lower ends of the four outer fuel assembly guide tubes are tapered gradually to form a region of reduced diameter which, in conjunction with the control element on the CEA, constitutes an effective hydraulic buffer for reducing the deceleration loads at the end of a trip stroke. This purely hydraulic damping action is augmented by a spring and plunger arrangement on the CEA spider.                When fully inserted, ' the CEAs and PSCEAs rest on the upper guide structure support plate.
              .The capability of the.B C CEAs to scram within the allowable time                                                        l was initially demonstraked as part of the flow tes' ting discussed                                                        I in Section                            4.2.4.4. The increased weight of the Ag-In-Cd                              y four-element CEA design also ensures that scram time limits will be met.                      Scram time has been verified through surveillance                                            I testing at the operating Palo Verde units.                                                                                l 4.2.3                                DESIGN EVALUATION
(
'                                                                                                                                        1 I
4.2.3.1                                Fuel Assembiv                                                                    l 4.2.3.1.1                                Vibration Analyses Four sources of external excitation are recognized in evaluating the fuel assembly susceptibility to vibration damage.                                                          These sources are as follows:                                                                                                    i l'            A.                Reactor Coolant Pump Blade Passing Frequency                                                            l Precritical vibration monitoring on previous C-E reactors
: i.                                indicates that peak pressure pulses are expected at the pump i
blade passing frequency (120 Hz), and a lesser but still                                              j pronounced peak at twice this frequency.                                                              i B.                  Lower Support Structure Motion Random lateral motion between the fuel assembly and the lower          support    structure    is          expected    to  occur  with            an      i kuplitude similar to that of other C-E reactors in the                                                l frequency range between 2 and 10 Hz.
1 Amendment F 4.2-44                            December 15, 1989 i
I
 
    . i
'I L            ; CESSAR W hi.
__C.'  Flow-induced fuel rod vibration results' from coolant flow through the fuel assembly. The expected amplitude of such
  +    '
vibration is 0.004 inches or less.
D. Flow Induced Control Element Vibration 0
System 80- incorporates - design features that minimize CEA vibration and produce no significant wear in the guide tubes.-
These sources of external excitation are not expected to have an adverse effect on - the performance _ of the fuel assembly.      The p capability of the- fuel assembly to sustain the effects of flow-induced    vibration  without    adverse effects  has- been
              -demonstrated in the dynamic flow tests reported in Appendix 4B.
4.2.3.1.2          CEA Guide Tube 0
The CEA guide tubes ware evaluated for structural adequacy using        t the criteria given in Section 4.2.1.1 in the following areas:
A. Steady axial load due to the combined effects of axial hydraulic forces and upper eid fitting holddown forces.
For normal operating conditions, the resultant guide tube.
stress levels are significantly less than the design limits.
B. Short-term axial load due to tle impact of the spring loaded CEA spider against the upper guide structure support plates at the end of a CEA trip.
For trips - occurring during normal power operation, solid impact is not predicted to occur due to the kinetic energy of the CEA being dissipated in the hydraulic buffer and by the CEA spring.
C. Short-term differential pressure load occurring in the
                    ' hydraulic buffer regions of the outer guide tubes at the end of each trip stroke.
The buffer region slows the CEA during the last few inches of the trip stroke. The resultant differential pressure across the guide tube in this region gives rise to circumferential stresses which are significantly less than the design limits. The trip is assumed to be repeated daily. However the resultant stress is too small to have a significant effect on fatigue usage.
l Amendment F        l 4.2-45              December 15, 1989
 
R    CESSAR iinhe,.
For    conditions  other than  normal  operation,  the  additional mechanical loads imposed on the fuel assembly by an OBE,-SSE, and        B    j large break LOCA and their resultant effect on the control                    ,
element guide tubes are discussed in the following sections:                  :
4.2.3.1.2.1        Operating Basis Earthquake (OBE)
During the postulated OBE, . the fuel assembly is subjected to lateral and axial accelerations which, in turn, cause the fuel
      = assembly to deflect from its normal shape.            The method of calculating these deflections is described in Section 3.7.3.14.
The magnitude of'the lateral deflections and resultant stresses
      -are evaluated for acceptability.        The method for calculating stresses from deflected shapes is described in Reference 56.        The        I fuel assembly is-designed to be capable of withstanding the axial              l loads without buckling and without sustaining excessive stresses.
g  4.2.3.1.2.2        Safe Shutdown Earthquake (SSE)
The axial and lateral loads and deformation sustained by the fuel assembly during a postulated SSE have the same origin as those discussed above for the OBE, but they arise from SSE initial            F ground accelerations. The analytical methods used for the SSE are identical to those used for the OBE.
4.2.3.1.2.3        Loss-of-Coolant Accident (LOCA)
In the event of a large break LOCA, there will occur rapid changes in pressure and flow within the reactor vessel.
Associated with the transient are relatively large axial and lateral loads on the fuel assemblies.        The response of a fuel              _'
assembly to the mechanical loads produced by a LOCA is considered acceptable if the fuel rods are maintained in a coolable array, i.e., acceptably low grid crushing.
4.2.3.1.2.4        Combined SSE and LOCA It is not considered appropriate to combine the stresses resulting from the SSE and LOCA events.            Nevertheless,    for purposes of demonstrating margin in the design, the maximum stress intensities for each individual event are combined by a square root of the sum of the squares (SRSS) method. This is
  ,    performed as a function of fuel assembly elevation and position, e.g., the maximum stress intensities for the center guide tube at the upper grid elevation (as determined in the analysis discussed in the above paragraphs for SSE and LOCA) are combined by the SRSS  method. Additional  details  regarding  the  analysis  of 3
combined seismic and LOCA loads are described in Reference 56.
I Amendment F 4.2-46                  December 15, 1989 1
 
h CESSAR !!nine.n .          ~
1 The results demonstrate that the allowable stresses described in Section 4.2.1.1'are not exceeded for any position along the fuel assembly even under the added conservatism provided by this load combination.
To qualify the complete                                      fuel  assembly,    full-scale hot        loop testing'was conducted prior to the initial operation of System 80                                                  F fuel. These tests evaluated fretting and wear of components, refueling procedures,                            fuel assembly uplift forces,                      holddown performance            and compatibility                            of  the    fuel    assembly    with interfacing reactor internals, CEAs and CEDMs under conditions of-reactor water                chemistry,                        flow    velocity,      temperature,      and pressure.            The  details        of          System    80  hot loop    testing    are  reported        B in Appendix 4B.
4.2.3.1.3                    Spacer Grid Evaluation The function of the spacer grids is to provide lateral support to fuel and burnable poison rods in such a manner that the axial forces are not sufficient to backle or bow the order and that the                                                            5 wear resulting at the grid-to-clad contact points will be limited                                                              - ,
to acceptably small amounts.                                      It is also a criterion that the grid be capable of withstanding the lateral loads imposed-during the postulated seismic and LOCA events.
Fuel assemblies are designed such that the combination of fuel rod rigidity, grid spacing, and grid preload will not result in significant fuel rod deformation under axial loads.                                          In addition, the long-term effects of clad creep (reduction in clad OD), the reduction of grid stiffness with temperature and the partial relaxation of the grid material during operation must be taken into account to ensure that this criterion is also satisfied during all operating conditions.                                      Inspections of irradiated fuel assemblies from the Maine Yankee (14 x 14), Calvert Cliffs (14 x                                                  p 14), Palisades (15 x 15) , Omaha (14 x 14), ANO-2 (16 x 16) , and Palo Verde (System 80) reactors has not shown significant bowing of the fuel rods.              In view of these factors, it is concluded that the axial forces applied by the grids on the cladding will not result in a significant degree of fuel rod bow. The influence of fuel rod lateral deflection is discussed further in Section 4.2.3.2.6.            Additional discussion of the causes for and effects of fuel rod bowing are contained in Section 4.2.3.2.5 and in Reference 57.
The capability of the grids to support the clad without excessive clad wear is demonstrated by out-of-pile flow testing on the Standard System 80 assembly design and by the results of post-irradiation examination of grid-to-clad contact points in extended burnup fuel assemblies in Calvert Cliffs, Omaha, and
;        ANO-2, as well as after two cycles of operation in Palo Verde                                                      p (System 80).
Amendment F 4.2-47                      December 15 1989
 
u L
CESSARiiih m.
          'The capability of                        the grid to withstand the    lateral    loads-      j produced                      during  the  postulated  seismic  and Inca events is demonstrated' by impact testing the reference grid design, and comparing the test results with the analytical predictions of.the seismic and Inca loads.
The Zircaloy-4 spacer grid material is of the same composition as the fuel rods ' and guide . tubes with which it is in contact,                                  ,
thereby eliminatirg any problem of chemical incompatibility with                          B t
those components.                        For the same reason, adequate resistance to corrosion from the coolant is assured (see Section                        4. 2. 3. 2. 3. ,      !
Item A, for additional information relative to the corrosion                                    '
resistance ' of Zircaloy-4 in the primary _ coolant environment).
The Inconal 625 material used for the lowest spacer grid is in contact with the coolant, the Type 304 stainless steel lower end fitting (to which it is welded), the Zircaloy-4 fuel rods, the poison rods,-and the Zircaloy-4 guide tubes. The mutual chemical compatibility of these materials in - a reactor environment has been demonstrated by C-E's use of these materials in fuel assemblies that have been operated in other C-E reactors and for which post-irradiation examination has yielded no evidence of chemical. reaction between these components.                            In. addition, experiments have been performed at C-E.on Inconel-type-alloys'and                            '
Zircaloy-4 which showed that eutectic reactions did not occur below 2200*F, a temperature far in excess of that anticipated at the lower grid location in the event of a IccA.
4.2.3.1.4                          Dimensional stability of Sircaloy Zircaloy components are designed to allow for dimensional changes resulting from irradiation-induced growth. Extensive analyses of in-pile growth                        data have    been performed to formulate a comprehensive model of in-pile growth (References 3 and 4). 'The in-pile growth equations are used to determine the minimum axial differential growth allowance which must be included in the axial gap between the fuel rods and the upper end fitting.                                  For determining: the gap between the fuel rods and the upper end fitting, the growth correlations for fuel rod and guide tube growth are combined statistically such that the minimum initial gap-is adequate to accommodate the upper 95% probability level of differential growth between fuel rods and guide tubes in the peak burnup fuel assembly.                        For the purpose of predicting axial and lateral                      growth    of  the  fuel  assembly structure    (thereby establishing the minimum initial clearance with interfacing components), the equations are used in a conservative manner to ensure adequate margins to interference are maintained.                              The manner in which the in-pile growth equations are used in design is described in References 4 and 58.
Amendment B 4.2-48              March 31, 1988
 
                                                                                                                                    ~
q l
l r            CESSAR !!!i%ui.                                                                                                          j I
l 4.2.3.1.5              Fuel Bandling and Shipping Design Loads l
Three specific design bases have been established for shipping .                                                            .
and handling loads. These are as follows.
                              ~
A.        The fuel assembly, when supported in the new fuel shipping
                      -container, shall be capable of sustaining the effect of Sg axial, lateral or vertical acceleration without sustaining stress levels in excess of those allowed for normal                                                              '
operation. The 5g criterion was originally established experimentally, and its adequacy is continually confirmed by the presence of impact recorders, as discussed in the following paragraph.
Impact recorders are included with each shipment which indicate if ' loadings in excess of 5g are sustained.                                                    A record of shipping loads in excess of 5g indicates an
;                        unusual shipping occurrence in which case the fuel assembly                                                  7 is inspected for damage prior to releasing it for use.                                                        A e  !
i
                      - The axial . shipping load path is through either end fitting                                                      e
                        'to the-guide tubes.- A Sg axial load produces a compressive
                      ' stress level in the - guide tubes less than the two-thirds                                                        -
        .                yield stress limit that is allowed for normal condition                                                          1 L:
events. The fuel assembly is prevented from buckling by                                                      -
being - clamped- at grid locations.                                For lateral or vertical                      ,
shipping loads, the grid spring tabs have an initial preload
!                        which exceeds five times the fuel rod weight.                                              Therefore,            f the spring tabs see no additional deflection as a result of l
5g lateral or vertical acceleration of the shipping container. In addition, the side load.on the grid faces produced by a Sg lateral or vertical acceleration is less                                                        .
than the measured impact strength of the grids.
B.      The fuel assembly shall be capable of sustaining a 5000-pound axial load applied at the upper end- fitting by the refueling grapple (and resisted by an equal load at the lower and fitting) without sustaining stress levels in excess of those allowed for normal operation.                                                    The 5000-pound load was chosen in order to provide adequate lift capability should an assembly become lodged.                                              This load i                        criterion is greater than any lift                                      load that has been encountered in service.
C.      The fuel assembly shall be capable of withstanding a 0.125-inch deflection in any direction whenever the fuel assembly is raised or lowered from or to a horizontal position without sustaining a permanent deformation beyond the fuel assembly inspection envelope.
4.2-49
 
lCESSARliiWican.                                                                        t Fuel handling procedures required the use of a strongback to limit. the fuel assembly deflection to a. maximum of 0.125-inch in any direction whenever the fuel assembly is raised or lowered to a horizontal position. This limits the                  !
stress and strain imposed upon the fuel assembly to values well below the limits set for normal operating conditions.                    )
The adequacy of the 0.125-inch criterion is based on the                    ;
inclusion      of .this  limitation    in    specifications    and        I procedures for fuel handling equipment, which is thereby constrained to provide support such that lateral deflection is-limited to 0.125-inches.                                                .
1
                          ^
4.2.3.1.6                  Fuel Assembly Analysis Results
:The results of the fuel assembly ~ analyses confirm that the design criteria' of Section 4.2.1.1, _regarding                  stress,    strain- and fatigue, are satisfied..                                                          f s
4.2.3.1.7                  Fuel A'ssembly Liftoff Analysis
          ~
The results of the analysis confirm that the fuel assembly will not experience liftoff during reactor operation. This analysis considers the appropriate combination of forces as described in Section 4.2.2.1.
4.2.3.2                Fuel Rod Desian Evaluation The evaluations discussed in this section are based on assumed fuel rod operation within certain linear heat rate limits relatad to avoiding excessive fuel and clad temperatures.                    Information concerning the bases for these limits is contained in Section 4.4.
4.2.3.2.1                  Results of Vibration Analyses B
Three sources of external excitation are recognized in evaluating L_    the fuel rod susceptibility to vibration damage.                  These sources are as described in Section 4.2.3.1.1.
These sources of periodic motion are not expected to have an adverse effect on the performance of the fuel rod.                        Section        J 4.2.3.2.4 includes additional information on fuel rod response to                        ,
the sources.                                                                            l l-    4.2.3.2.2                  Fuel Rod Internal Pressure and Stress Analysis i
A fuel rod cladding stress analysis is conducted to determine the                        I circumferential stress and strain resulting from normal, upsat, and emergency conditions. The analysis includes the calculation                          l of cladding temperatures and rod internal pressures during each of the occurrences listed in Section 4.2.1.1.                        The design Amendment F 4.2-50                  December 15, 1989
 
iu C E S S A R Rn Wicui.                                                                i e
criteria to be used to evaluate the analytical results are specified in Section 4.2.1.2.1. Fuel rod stresses resulting from seismic events are calculated, using the methodology described in Reference 56.
The results of the fuel rod analyses confirm that the design criteria of Section 4.2.1.2.1,-          regarding- stress,    strain and  p strain fatigue, are satisfied.
4.2.3.2.3          Potential for Chemical Reaction                                y A. Corrosion                                                                      )
Zircaloy-4 fuel rod tubing has been visually examined in the                    ;
spent fuel pool after four reactor cycles at Ft. Calhoun, four reactor cycles at Calvert Cliffs, and others. at Millstone 2, St. Lucie-1, and Maine Yankee.            In addition, oxide thicknesses were measured in the hot cell after one cycle at Maine Yankee.        The oxide appearance and thickness          i measured were similar to that from autoclave behavior for that time and temperature.                                                $      1 Coolant chemistry parameters have been specified that                            '
minimize corrosion product release rates and their mobility                      1 in the primary system.          Specifically, the pre-core hot functional environment is pH and oxygen controlled to                            ;
provide a thin, tenacious, adherent, protective oxide film.                        !
This approach minimizes corrosion- product release and                            I associated inventory on initial startup and subsequent operation.      During  operation,  the  recommended    lithium concentration      range (1.0-2.0 ppm) effects a chemical potential gradient or driving force between hotter and cooler surfaces (fuel cladding and steam generator tubing, respectively) such that soluble iron and nickel species will preferentially deposit on the steam generator surfaces.          The associated pH also minimizes general corrosion product                              l release rates from primary system surfaces. Morawer, the specified hydrogen concentration range ensures reducing cogitions in the core,          thereby avoiding low solubility Fe  . Additionally,      dissolved h}drogen promotes rapid recombination of oxidizing species.            (Recall,  oxidizing
              -species    and      a  fast  neutron    flux    are    synergistic prerequisites to accelerated Zircaloy-4 corrosion).
During operation, lithium, dissolved oxygen, and dissolved hydrogen will be monitored at a frequency consistent with maintaining these parameters within their specifications, as identified in Table 9.3.4-1.                                          F Amendment F 4.2-51                December 15, 1989
 
,  TCESSAR timne m.
Post-operational examinations of fuel- cladding that has operated  within      these  specifications    have  shown    no significant chemical or corrosive attack of the Zircaloy cladding.
B. External Hydriding                                                      ,
During operation of the reactor, with exposure to high temperature, high pressure water, Zircaloy-4 cladding will              '
react to form a protective oxide film in accordance with the following equation:
Zr + 2H 2 O
* ZrO2 + 2H2                                        I Approximately 20% of the hydrogen is absorbed by              the Zircaloy. Based on data described in WAPD-MRP-107,          the cladding would be expected to contain up to 250 ppm of hydrogen following 3 years of exposure.                                    i A series of burst tests were performed on Zircaloy-2 tubes                ;
containing 340 ppm and 460 ppm of hydrogen precipitated as                  '
hydride platelets in a ci cumferential manner (Reference 59). Burs' tests at 6 64 F showed -that the burst test specimens vith 340 ppm hat normal burst ductility of .12%.
Therefore, hydrogen normalif absorbed in Zircaloy-4 tubing I
will not compromise claddin, integrity.
C. Internal Hydriding                                                          :
A number of reported fuel rod failures have resulted from excessive moisture available in the fuel. Under operation, this moisture oxidizes the Zircaloy. -A fraction of the            B hydrogen, which is generated during normal oxidation, would be absorbed into the Zircaloy.          This localized hydrogen absorption by the cladding would shortly result in localized fuel rod failure.
Work performed at the OECD Reactor Project, Halden, Norway, of which C-E is a member, demonstrated that a threshold value of water moisture is required for hyfride sunbursts to occur (Reference 60).          Through  a  series  of  in-pile              i experiments,    the  level  of  this  threshold  value    was established. The allowable hydrogen limit in the fuel complies with this requirement,          ensuring that hydride sunbursts will not occur.
D. Crud Crud layers on zirconium oxide films are usually porous and non-insulating.      As an example, heavy, but non-insulating crud layers have been found in Yankee Rowe (WCAP-3317-6094, Amendment B 4.2-52                March 31, 1988
 
C E S S A R BM Wie. m r Yankee Core Evaluation Program, Final Report, 1971). .With porous crud, water , is free to flow through the crud and-provide    heat    transfer        by  convection. Under  these conditions, crud enhanced corrosion should not occur.
Because of rigorous water chemistry monitoring, aeavy buildup of crud has not occurred in C-E reactors.              Water chemistry monitoring is a continuous process and should ensure no' dense crud buildup.
E. Fuel-Cladding Chemical Reaction An in-depth post-irradiation examination has been conducted wherein fuel-cladding chemical reactions were among those items studied (Reference 26). .This study concluded that early unpressurized elenents containing unstable fuel were more susceptible to stress corrosion attack than are the current design that utilizes stable fuel and pressurized                F cladding. Since stress corrosion attack is the result of a                y combination of stress imposed by the fuel on the cladding                    i and the corrosive chemical species available to the cladding, irradiation programs have been pursued to define the conditions under which pellet-clad interaction will                  B damage the cladding. These programs have been conducted at                  ,
Halden, at Petten in the Netherlands, and at Studsvik in Sweden, and have confirmed that current fuel designs are not susceptible to failure by stress corrosion cracking during normal plant operation.
4.2.3.2.4        Fretting corrosion The phenomenon of fretting corrosion, particularly in Zircaloy clad fuel rods supported by Zircaloy spacer grids, has been extensively    investigated.      Since irradiation-induced stress
: relaxation causes a reduction in grid spring load, spacer grids must be designed for end-of-life conditions                  as well    as beginning-of-life conditions to prevent fretting caused by flow induced tube vibrations.
Examination of Zircaloy clad fuel rods after six cycles of                    B exposure at Ft.      Calhoun, five cycles at Calvert Cliffs-1, and,F five cycles at ANO-2 indicate fuel rod fretting between the fuel rod and spacer grid is rare.              The usual result of the contact between grid components and fuel rods is a small cladding surface B mark with no appreciable depth.
4.2.3.2.5          Fuel Rod Bowing Experience has proven that any specific criterion on allowable deflections (bowing), with respect to the effects which such deflections might have on thermal-hydraulic performance, is not Amendment F 4.2-53                    December 15 1989
 
u
      ;CESSAR !!!Mem.
1 necessary beyond the initial fuel rod positioning requirements                                    I required of the grids. This variation in spacing is accounted for-in-thermal-hydraulic analysis through the introduction of hot channel factors in calculating the maximum enthalpy rise in calculating DNBR.        This adjustment is called the Pitch, Bowing, and clad ' Diameter 2nthalpy- Rise Factor, which is conservatively applied to simulate a reduced flow area along the entire channel length.        The value of this factor is given in Table 4.4-1 and its application is discussed in section 4.4.
The subject of fuel rod bowing is discussed in Reference 57, 4.2.3.2.6            Irradiation Stability of Fuel Rod Cladding The combined effects of fast flux and cladding temperature are considered in three ways as discussed below:
A.        Cladding Creep Rate-The in-pile creep performance of Zircaloy-4 is dependent upon both the local material temperature and the local fast
;                  neutron flux. The functional form of the dependencies is presented    in        References                15-17    for  gap    conductance calculations, and in Reference 25 for cladding collapse time predictions.
;        B.        Cladding Mechanical Properties L                  The yield strength, ultimate strength, and ductility of l
Zircaloy-4 are dependent upon temperature and accumulated                              -
fast    neutron      fluence.                  The    temperature    and    fluence dependence    are                    discussed    in    Section    4.2.1.2.2.1.
Unirradiated properties were -used depending upon which is more restrictive for the phenomenon evaluated.
j        C.        Irradiation Induced Dimensional Changes Zircaloy-4 has been shown to sustain dimensional changes (in the unstressed condition) as a function of the accumulated fast  fluence.              These changes              are  considered in the appropriate clearances between the various core components.
The irradiation induced growth correlation method is discussed in Reference 3.
Zircaloy-4 fuel cladding has been utilized in pressurized water reactors at temperatures and burnups                                anticipated in current designs with no failures attributable to radiation damage.
.        Mechanical property tests on grealcy-4 cladding exposed to                                      .
i neutron irradiation of 4.7 x 10                              nyt (E>l MeV) (estimated) have i
1 4.2-54
  -.. ~.    .. . .                  . _ , _ . - . . . - .
 
y i
W    CESSAR !!!Ln.
revealed that the cladding retains a significant amount of ductility (in excess of 4% elongation).        Typical results are shgnnytin Table 4.2-2.
(E>l MeV) is It is believed that the fluence of 4.7 x at satoration so that continued exposure to 10 irradiation will not change these properties (Reference 61).
4.2.3.2.7        Cladding Collapse Analysis A cladding collapse analysis is performed to ensure that no fuel rod in the cora will collapse during its design lifetime. The clad collapse calculation method (Reference 24) itself does not include arbitrary safety factors.        However, the calculation      p inputs are selected to produce a conservative result.            For example, -the clad dimensional data are chosen to be worst case combinations based either upon drawing tolerances or 95%
confidence limits on as-built dimensions; the internal prossure history is based on minimum fill pressure with no assistance from released fission gas; and the flux and temperature histories are based on conservative assumptions.
4.2.3.2.8        Fuel Dimensional stability Fuel swelling due to irradiation (accumulation of solid and gaseous fission prodpcts) and thermal expansion results in an increase in the fuel pellet diameter. 'The design makes provision for accommodating both forms of pellet growth.        The fuel-clad diametral gap is more than sufficient to accommodate the thermal expansion of the fuel.        To accommodate irradiation-induced swelling, it is conservatively assumed that the fuel-clad gap is used up by the-thermal expansion and that only the fuel porosity and the dishes on each end of the pellets are available. Thermal and irradiation induced creep of the restrained fuel results in redistribution of fuel so that the swelling due to irradiation is accommodated by the free volume (8.2% of the fuel volume).
For such restrained pellets, and at a 2p      tal figsion-product-    B induced swelling rate of 0.4% AV/V per 10    fiss/cm , 0.24% would be accommodated by the fuel porosity and dishes through fuel creep, and 0.16% would increase the fuel diameter. Assuming peak burnup, this would correspond to using up a void volume equal to 3.3% of the fuel volume and increasing the fuel rod diameter by a      B maximum of <0.0025 inch (<0.7% clad strain) . When these numbers were compared to the minimum available volume and the maximum allowable strain, it was concluded that sufficient accommodation volume has been provided even under the most adverse burnup and tolerance conditions.
i Amendment F 4.2-55                December 15 1989
 
l CESSARnia m,.                                                                    !
Early work on the            swelling ratio for UO      is described    in References 8 and 62 through 65. These experi$ents were conducted                )
using fuel materials made prior to the discovery of densification and would be appropriate-for some of C-E's early production.          The        I incorporation of pore formers that provide more representative                    !
fuel microstructures makes a more recent set of data from a C-E                  I conducted program more appropriate for swelling values.
Fuel pellets were fabricated by C-E and irradiated in Calvert Cliffs for four cycles to burnup levels up to 50,000 mwd /MTU                B (Reference      66). Immersion    density measurements taken from pellets of various burnups were plotted to determine the rate of volume increase with burnup.              The rate derived 'from these measurements is 1.0% AV/V per 10,000 mwd /MTU.          Since the levels of burnup were after cladding contact, the swelling value obtained is a restrained swelling rate. The 0.4% value per 4,0g                  .
mwd /MTU  is approximately equal to the 0.4% AV/V per                10          l 3
fissions /cm used for the calculations described above.
4.2.3.2.9            Potential for Waterlogging Rupture and Chemical Interaction The potential -for waterlogging rupture is considered remote.
Basically, the necessary factors, or combination of factors, include the presence of a small opening in the cladding, time to permit filling of the fuel rod with water, and finally, a rapid
^
power transient.        The size of the opening necessary to cause a problem falls within a fairly narrow band.              Above a certain defect size, the rod can fill rapidly, but during a power increase it also expels water or steam readily without a large pressure buildup.        Defects which could result in an opening in cladding are scrupulously checked for during the fuel rod manufacturing process by both ultrasonic and helium leak testing.
Clad defects which could develop during reactor operation due to hydriding are also controlled by limiting those factors; e.g.,
hydrogen content of fuel pellets, which contributes to hydriding.
The most likely time for a waterlogging rupture incident would be after an abnormally long shutdown period.              After this time, however, the startup rate is controlled so that even if a fuel rod were filled with coolant,              it would " bake out",    thus minimizing the possibility of additional cladding rupture.            The combination of control and inspection during the manufacturing process and the limits on the rate of power change restrict the potential for waterlogging rupture to a very small number of fuel rods.
Amendment B 4.2-56              March 31, 1988
 
z                                                                                                    1 CESSAR !!nLn .
The UO      fuel pellets are highly resistant to attack by reactor                              i cooland in the event cladding defects should occur. Extensive experiaantal work and operating ,sx;perience have shown that the design parametery chosen conservat;,vely account for changes in thermal performance during operation and that coolant activity buildup resulting from cladding rupture is limited by the ability of uranium dioxide to retain solid and gaseous fission products.
4.2.3.2.10          Fuel Burnup Experience The C-E fuel rod design is based on an extensive experimental data base and by an extencion of experimental . knowledge through design application of C-E fuel rod evaluation codes.                                  The      i experimental data base includes data from C-E or C-E/Kraftwsrk Union (KWU) joint irradiation experiments,. from . C-E and KWU operating commercial plant performance and from many basic experiments conducted in various research reactors which are available in the open literature.                        Some of these sources are          #
discussed - below. Evidence            currently          available indicates that Zircaloy and Uo., fuel performance is satisfactory to exposures in                            ,
l excess of 55,000 mwd /MTU.
A. Public Information General fuel perform'ance information available in th'e open literature has provided ;part of the C-E fuel rod design data base.      Particular exper;.ments that have been cited in the                          7, past as key references are:                                                              ;
: 1.      Determination of the effect of fuel-cladding gap on the linear heat rating to melting                        for UO            rods, l
conducted in the Westinghouse test reacto$. fuel
: 2.      Shippingport Irradiation Experience.
: 3.      Saxton Irradiation Experience.
: 4.      Combined    Vallecitos                    Boiling        Water    Reactor (VBWR)/Dresden irradiation.
: 5.      Large Seed Blanket Reactor (LSBR) Rod Experience.
: 6.      Joint U.S.-Euratom Research and Development Program to evaluate central fuel melting in the Consumers Power Co. Big Rock Point Reactor.
l Since the information from these programs is available in the open literaturc, they will not be described here.
However, details as to the significance of the results to C-E fuel burnup experience are presented in Reference 67.
l l
4.2-57
 
z CESSAR1!EL m.,
B. C-E/KWU Technical Exchange                                                  I C-E entered into a technical agreement with KWU beginning in-1972 for the ccmplete exchange of information and technology relating to pressurized water reacter systems, including fuel. This agreenent made available to C-E the total-                    !
o        experience of 10 years successful . operation of commercial PWR ' fuel . in systems ' designed and fabricated 'by KWU and is            !
the most advanced of- its type in the world.
An essential part    of      this broad-based        exchange  involved      joints sponsorship'of numerous-fuel testing programs.
C. Operating Fuel Experience                                                    l C-E has fabricated mere than 1,500,000 Zircaloy-clad fuel rods both internally pressurized and unprescurized. of this total, 545,000 rods remain in operation, some with average              F burnups in excess of 50,000 mwd /MTU. Overall performance of this fuel has been excellent.            The fuel rod reliability level,. satimated    from coolant activities,            is' >99.99%.
Reliability levels are continually validated by extensive poolside fuel inspection programs conducted by C-E at                  g reactor sites during refueling shutdowns.
D. Fuel Irradiation Programs                                                    i C-E is involved      in diversified        fuel irradiation test programs to confirm the adequacy of the C-E fuel rod design bases. and models by experimental means.            Some of these programs involve safety related .research while other programs provide confirmatory data on performance capability or evaluate design and fabrication variables or methods which may improve and extend our current knowledge of rod performance.
fuel lB Some of the key fuel performance evaluation programs that are summarized below include:                                          B Fuel densification experiments at the Battelle Research Reactor (BRR);
Joint C-E/KWU fuel densification experiments including tests in the MZFR at Karlsruhe,          West Germany, and the EEI experiments in the General Electric Test Reactor (GETR);
Direct participation in the Halden Project in Norway with access to all Halden base program fuel test data; Amendment F 4.2-58                  December 15, 1989          j
 
e i
[
JCESSAR!R L ..o Irradiation of special instrumented fuel rods to obtain g                dynamic in-reactor measurements in Halden experimental rigs; Ramp t'est programs on fuel rods to' evaluate fuel load-follow capabilities                  and        .the    pellet-clad    interaction / stress 1              corrosion phenomenon in both the studsvik and Petten test reactors, and-other in-reactor experiments conducted in the obrigheim-pressurized water reactor; and, Irradiation of special test and surveillance assemblies in operating C-E reactors.
E. C-E Fuel Densification Experiments h                C-E has conducted several experiments which provided data on the in-reactor- densification behavior of various Uo, fuel types. These-include the BRR, MZFR, and EEI densifft:ation experiments,-which are described below.                                                  .y; .
_            F. BRR Fuel Densification Experiment The object of~ this program was . to examine the                                in-pile densification                        behavior. *of      various    fuel  types    and microstructures fabricated ' with and without pore formers.
g                The non-pore former fuel-types'had initial densities of'93%
L                  to 94% theoretical with a grain size of less than 6 microns, 5                  with a larga fraction of pores less than 4 microns in                                      ,
_                  diameter. The porc former fuel types had initial densities                              e of 93% to 95% and were characterized ' by a combination of                                +
large grain size and/or large pore size.                              Fuel pellets of each experimental type were irradiated in six BRR capsules at linear heat ratings between 2.8 and 4.6 kw/ft for periods of up to 1500 hours.                          Post-irradiation examination of the BRR    results                      showed    significant    differences    in    the densification behavior between pore former and non-pore former fuel.                The pore former fuel 'showed little change in density (high stability) while the non-pore former fuel densified rapidly. A trend towards increased densification with lower initial density was apparent in the non-pore former' fuel.                It was concluded that the UO microstructure played a dominant role in the kinetics2 and extent of hi-reactor densification. Consequently, fuel exhibiting the desirable microstructural features to reduce in-reactor densification (i.e., large fraction of the pore volume in the large pore size range) became part of the standard C-E fuel design.
G. C-E/KWU Fuel Densification Experiment (MZFR)
As a follow-on to the C-E experiment in the BRR, a joint C-E/KWU program has been conducted in the German MZFR to 4.2-59
 
CESSARlinL e 1
evaluate the performance of several non-densifying                        fuel    .
types at higher power levels for longer times and to higher t                  burnups, o
Sixteen full-length fuel rods each containing a different                            -
fuel type were irradiated at powers up to il kW/ft for                              -
burnups up to-4000 mwd /MTU.        Included.in these rods are UO fuels most of which were fabricated usinh techniqliesand - UO -Puo$ntended to minimize densification.              Six rods employed C-E fabricated UO 2 fuels, five of which included pore former additives and one fabricated without a pore former to serve as ' a referenceable control sample.                    Eight M            '
rods    were    fabricated    using        KWU    experimental          fuel l                  representing      a  wide range      of      sintering. times and temperatures,      initial    densities      and enrichments.            The    ,
remaining two rods were fabricated using UO 2 -Puo fuels of two different densities, with and without a pdre former additive.      Each of the fuel pellet types and fuel rods was
;.      '          extensively characterized prior to testing to. permit 1          comparison with similar post-irradiation measurements.
The results of the post-irradiation examination showed that fuel types fabricated with pore formers (similar_to current                        I J                  production fuel) experienced: significantly less in-pile                            ,
densification- compared to those fabricated without pore formers.      The _ data also support use of a. standardized out-of-pile      resintering    test        developed- by        C-E      to
'                  characterize expected in-pile densification at the time of fabrication. This simulation test has been submitted to the NRC and approved for use by C-E in LOCA calculations.
H. EEI Fuel Densification Experiment F      '
The prime objective of the EEI Fuel Irradiation Test Program conducted in the General' Electric Test Reactor (GETR) was to isolate and        characterize    the      in-reactor      densification-behavior of pore former (or stable) fuel. types. C-E and KWU were among eleven participants in the program.
This program entitled C-E to obtain densification data on
(-                  nine    base    program    fuel    pellet      types      with      varying
.                  microstructures.        An additional          four    fuel    types were L                    fabricated by C-E and KWU.        These included C-E fuel types, two with and one without a pore former additive and a KWU standard production fuel. The pellets in the program were well characterized prior to irradiation.                Four of the fuel types were irradiated in one pressurized (53 atmospheres) capsule. Two of the fuel types were also irradiated in a
<                  separate non-pressurized capsule (one atmosphere).                    Each of the capsules contained thermocouples to continuously monitor e
4.2-60
 
CESSARHnLm.
capsula power generation during irradiation to assure that the desired operating conditions were maintained.        Post-irradiation examination of these test capsules confirmed that UO 2 fuel with specific microstructural characteristics, such as produced by pore former additives, are stable with respect to densification. The largest in reactor density changes occurred for those types having a combination of the smallest pore size, the largest volume percent of porosity
(  less than 4 microns) in the smallest initial grain size and the lowest initial density (Reference 68).        .,
I. Halden Program Participation The experimental facilities and programs of the OECD Reactor Project in Halden, Norway represent one of the most advanced efforts in quantifying the effects and interaction of the various design parameters of Zircaloy-clad fuel rods through measurements made in reactor. C-E has been a member of the Project since 1973. C-E reviews the data generated by the project in considerable detail and utilizes the results in various fuel development programs.
The Halden test reactor has unique capability for measuring fuel rod operation during irradiation. This capability has been utilized by C-E with specific experiments to provide information in the following areas:
Fuel densification phenomena including measurements of the rate of fuel column shortening as a function of the initial fuel density, power level and fuel fabrication process.
Fuel clad mechanical interaction involving studies of the effects of pellet design (shape and density) and operating parameters on cladding deformation.
Modeling of fuel rod behavior with emphasis on heat transfer characteristics.
The first three test assemblies sponsored jointly by C-E and KWU contained 24 well-characterized      fuel  rods. These asseml:> lies included the following range of design and operating parameters:
Helium fill pressures from 22 to 35 atmospheres; Initial fuel densitics from 91-96% TD; Linear heat ratings to 15 kW/ft; and, 4.2-61
 
              . CESSAR !!nincam                                                                        .
:L U      enrichments from 6 to 12wt%. (9 rods fabricated with
                        -m$Nd-oxide fuel).
The' objectives of these tests were to determine the dynamic changes in fuel rod internal pressure, fuel centerline temperature and fuel stack length during operation as a function of burnup.        Two of these assemblies (6 test rods each) were discharged from the reactor- after receiving a                      '
                        . peak burnup'of 24,000 mwd /MTU.          The third rig' (12 rods) was irradiated to a peak burnup of 40,000. mwd /MTU so that fuel swelling and gas release behavior could be evaluated to high                B burnups. The objectives of a fourth six-rod test assembly
                        . ware to evaluate the effects of such design variables as pellet-clad gap, fill-gas composition, and linear heat rating.(to 15 kw/ft) on heat transfer characteristics. This experiment also provided gap conductance data on UO and mixed-oxide. fuel.      This test was discharged from the rd ctor after reaching a peak burnup of 4000 mwd /MTU.
7,                Instrumentation used to measure                fue_1  behavior during irradiation includes centerline thermocouples,                      internal pressure      transducers,          linear    variable      differential transformers (LVDTs) for fuel column length changes and flux monitors for axial and radial power profiles.
Fuel    column    length change data obtained support data                  8 l
      ~
generated by the EEI, BRR, and MZFR experiments and confirm the in-reactor stability of C-E pore former fuel types.                  In addition, the internal pressure monitors and centerline thermocouple data have confirmed the adequacy of the C-E thermal performance design models.
In addition to these C-E/KWU test assemblies, C-E has designed and irradiated three rods in the Halden high temperature, high pressure loop to simulate PWR coolant temperature and pressure conditions. The purpose of these B
experiments    was    to    distinguish    the  effects      of  pellet configuration on the formation of circumferential ridging and on the elongation of the rods. Each rod contained three pellet types with one type as a standard.              This program in combination with the results of other experiments gives C-E a firm basis upon which to optimize fuel rod design with respect to      dimensional changes and to improve fuel performance models developed to predict rod dimensional stability.
Amendment B 4.2-62                  March 31, 1988
 
CESSAR !!nMeam J.- -Power Ramp Programs C-E      and _ .KWU    participated                                          in    the    Studsvik    and Pathfinder /Petten programs to evaluate fuel rod -performance under ramp conditions to power levels not recently attained.
These can occur either after refueling or after extended periods of low power operation or during control rod
                . maneuvers. The effects of various fuel rod design variables on power ramp limits are also investigated as a means to-further optimize design.                                        The Petten/ Pathfinder program which began in 1973 is being- conducted jointly by C-E and KWU in the obrigheim PWR reactor and Petten test reactor facilities (Reference 69). One special test assembly has                                                    g been irradiated each year from 1973 to 1980 in the Obrigheim reactor.      Included in this assembly, which is designed to facilitate      fuel  rod            removal                                  and      replacement,    are well-characterized segmented rods or "rodlets" which are axially connected to form a complete fuel rod.                                                        These    a rodlets were " pre-irradiated" in the Obrigheim reactor for one to four operating cycles, and then separated and                                                            n irradiated in a test reactor te evaluate performance under ramp conditions. .Ninety-nine of these rodlets irradiated in
                'Obrigheim have been discharged and ramped - in Petten.                                                    An B additional 40 of these rodlets have been tested at the R-2 reactor at Studsvik.            Post-irradiation, hot-cell examination programs form an integral part of both the Petten/ Pathfinder                                                    ..
and Studsvik experiments to characterize fuel rod behavior, particularly with respect to dimensional stability and                                                          e fission product release.                      These test programs are designed                                y to distinguish between fuel rod power ramps which occur on start-up and those which might occur during reactor power maneuvering operations.
Plant operating flexibility requires that the fuel rods maintain integrity during periodic changes in power.                                                  Power cycling tests of this type have been jointly conducted by C-E/KWU in Obrigheim and Petten.                                                In the Petten test, a single unpressurized fuel rod was power cycled between 9 kW/ft and 17 kW/ft at a power change rate of about 3 kW/ft/ min. The fuel rod successfully completed 400 cycles and achieved a burnup of 8000 mwd /MTU. Power cycling tests were then conducted in Obrigheim on eight short pressurized and unpressurized fuel rods.                                                The test fuel rods were attached to a control rod drive mechanism and driven from the low power to a high power position on a nominal cycle.
Power changes from 50% to 100% at rates of 20% per minute for 880 cycles were included. After successfully completing the experiment, the test rods achieved a peak burnup of 30,000 mwd /MTU without substantial cladding deformation or fuel rod perforation.
Amendment B 4.2-63                                                  March 31, 1988
 
jb LC E S S A R !!n h o ,.
L l                                                                                                            J l
l '-
K. Fuel Surveillance Programs I
C-E  hasoperating conducted                a numberThusoffar, fuel surveillance  programs on lB fuel'in                          plants.              a total of thirty-eight poolside fuel inspection programs of varying detail have been performed by C-E (see Table 4.2-3) .                    A large number of lB assemblies have been visually examined, and dimensional.                                  1 measurements have been obtained on a large number of these assemblies. Fuel bundle disassembly: -operations have been-conducted either .to obtain                      information on particular performance aspects'or as part of test assembly surveillance-programs. A listing of these-programs and a summary of the                        B results is provided in-Reference 70. The results of the C-E poolside'' inspection program have been used to verify fuel assembly operation and provide data in support of design.                        A L                    poolside fuel surveillance program is being conducted at                          B Palo Verde-1 for C-E's System 80 fuel (see Section
: 4. 2.1. 5.1) .
4.2.3.2.11                    Temperature Transient Effects Analysis 4.2.3.2.11.1                      Waterlogged Fuel                                          j l
The potential for- a fuel rod to become waterlogged during normal operation is discussed in section 4.2.3.2.9. In the event that a fuel rod does become waterlogged at low or zero power, it is possible that a ' subsequent power increase could cause a buildup                            .I of hydrostatic pressure.                  It is unlikely that the pressure would build up to-.a level that could cause' cladding rupture because a fuel pin with the potential for rupture requires the combination of a very small defect together with a long period of operation at low or zero power.                                                                            )
7 Tests which have been conducted using intentionally waterlogged fuel pins (capsule drive core at SPERT) (References 71 and 72)                                    ,
showed. that the resulting failures -did eject some fuel material
,.              from the rod and greatly deformed the test specimens.                        However,            l l              these test rods were completely sealed, and the transient rates                                  l l>              used were'several orders of magnitude greater than those allowed L              in normal operation.
In those instances where waterlogged fuel rods have been observed in commercial reactors, it has not been clear that waterlogging was the cause, and not just the result, of associated cladding failures; and C-E has not observed and is not aware of any case in which material was expelled from waterlogged fuel rods or in
              .which the fuel cladding was significantly deformed in a normal power reactor.
Amendment B 4.2-64                  March 31, 1988              l
 
CESSAR tinhi.
- It ' is therefore . concluded that the effect of normal power
' transients on waterlogged fuel rods is.not likely- to result in cladding ' rupture and - even if rupture' does occur ~ it will 'not produce the sort of postulated burst failures which would expel
-fuel. material or damage adjacent fuel rods or fuel assembly structural components.
4.2.3.2.11.2:        Intact Fue1~
The . thermal effects of anticipated operational- occurrences on fuel red integrity are discussed in the following paragraphs.
A. Fuel rod thermal transient effects are basically manifested as the change in internal prassure, the changes in clad thermal gradient r.nd thermal stresses, and the differential thermal _ expansion between pellets and clad. These effects are discussed in Sections 4.2.3.2.2 and 4.2.3.2.11.
B. Another .possible effect of transients would be to cause an
* axial- expansion of the pellet column against a flattened (collapsed) section of the clad.      However, the fuel rod design  includes    specific provisions    to  prevent clad flattening, and, therefore,        such interactions will not occur.
4.2.3.2.12        Energy Release During Fuel Element Burnout s
The reactor protective system provides fuel clad protection so        "
e that the probability of fuel element burnout during normal operation and anticipated operational occurrences is extremely low. Thus, the potential for fuel element burnout is restricted to faulted conditions. The 14CA is the limiting event since it results in the larger number of fuel rods experiencing burnout; thus, the LOCA analysis, which is very conservative in predicting fuel element burnout,-provides an upper limit for evaluating the consequences of burnout.      The IDCA analysis explicity accounts for - the additional heat release due to the chemical reaction between the Zircaloy clad and the coolant following fuel element burnout in evaluating the consequences of this accident.        LOCA analysis results are discussed in Section 15.6.5.
4.2.3.2.13        Energy Release on Rupture of Waterlogged Fuel            ;
Elements                                                i A discussion of the potential for waterlogging of fuel rods and for subsequent energy release is presented in Section 4.2.3.2.10.
4.2-65
 
a CESSAR !!iWicm.s                                                                I l
I 4.2.3.2.14        Fuel Rod. Behavior Effects from Coolant Flow Blockage-An experimental and analytical program was conducted to determine                i the effects of fuel assembly coolant flow maldistribution during                l norms 1 reactor operation.
In the experimental phase, velocity and-static pressure measurements were made in cold, flowing water              J in an oversize model of a C-E 14 x 14 fue?. assembly in order to              '
destermine the three-dimensional        flow distributions    in  the          l vicinity of several types of flow obstruction. The effects of                1 the distributions on thermal behavior were evaluated, where                      l necessary, ~ with the use of a preliminary version of the TORC              .I thermal and hydraulic code (Reference 73).                                      l s
Subjects investigated included:
i A. The assembly . inlet flow maldistribution caused by blockage of a core support plate flow hole. Evaluation of the flow recovery data indicated that even the complete blockage of a core support plate. flow hole would not produce a.W-3 DNBR of less than 1.0 aven though the reactor might be operating at a power sufficient- to produce a . DNBR of 1.3 without the blockage.
B. The flow maldistribution within the assembly caused by complete  blockage    of    one  to 'nine  channels. Flow
          . distributions were measured at positions upstream and downstream of a blockage of one to nine channels.            The influence of the blockage diminished very rapidly in the upstream direction.      Analysis of the data for a single channel blockage indicated that such a blockage would not produce a W DNBR of less than 1.0 downstream of the                    l blockage even though the reactor might be operating at a power sufficient to produce a DN8R of 1.3 without the blockage.
The results presented above were obtained through flow testing an oversized model of a standard 14 x 14 fuel assembly.        Because of the great-similarity in design between the Standard System 80 16 x 16 assembly, and the earlier 14 x 14 array, these test results also constitute an adequate demonstration of the effects that flow blockage would have          on the 16 x 16 assembly.        This conclusion is also supported by the fact that the 16 x 16 assembly has been demonstrated to have a greater resistance to axial flow than would occur with the 14 x 14 array.        The effect of the higher flow resistance, to produce more rapid flow recovery, i.e., more nearly uniform flow, is analogous to the common use of flow resistance devices (screens or perforated plates) to smooth non-uniform velocity profiles in ducts or process equipment.
4.2-66
 
l I
CESSARRBL                                                                                                                                    ;
I i
4.R.3.3.15                          Puol Temperatures Steady rstate              fuel temperatures are determined by the FATES computer program.                      The calculational procedure considers the                                                            i effects of linear heat rate, fuel relocation, fuel swelling,                                                                                  I densification, thermal expansion, fission gas release, and clad                                                                              !
deformation. The model for predicting fuel thermal performance in. discussed in detail in References 15-17.
Two sets                of burnup and axially dependent linear heat rate distributions are considered in the calculation.                                                                One is the hot rod,- time averaged, distribution expected to persist during long-term operation, and the other is the envelope of the naximum linear heat rate at each axisl                                    location.                                      The long-term distributions are integrated over selected time periods to                                                                                    .
determine burnup, which are in turn used for the var:,ous burnup dependent behavioral models in the FATES computer program.                                                                          The  .. :
envelope accounts for possible variations in the reak linear heat                                                                        a ;
rate at any elevation which may occur for short periods of time                                                                          3  -
and is used exclusively for fission gas release calculations.
The power history used assumes continuous 100% reactor power from                                                                            -
beginning-of-life.                      Using thi;- history, the highest fuel temperatures occur at that time.                                It has been shown that fuel                                                i temperatures              for a given                  power level at any burnup are                                                    t insensitive to the previous history used to arrive at the given                                                                            L power level.
l Tual thermal performance parameters are calculated for the hot rod.              These parameters for any other rod in the core can be obtained by using the axial location in the hot rod, whose local power and burnup corresponds to the local power and burnup in the rod being examined.                        This procedure will yield conservatively high stored energy in the fuel rod under consideration.                                                                                      ,
The maximum power density, including the local peaking as affected by anticipated operational occurrences, is discussed in Sections 4.3, 4.4, and Chapter 15.
4.3.3.3                    Burnable Poison Red 4.3.3.3.1                      Burnable Poison Rod Internal Pressure and                                                                    ,
  ;                                        Cladding stress The poison rod cladding will be analyzed to determine the stress L          and strain resulting from the various normal, upset,                                                                                and emergency conditions discussed in Section 4.2.1.1.                                                                              Specific
[          accounting is made for differential pressure, differential i
I thermal expansion, 01 adding creep, and irradiation induced swelling of the burnable poison naterial.
4.2-67
 
l CESSAR tminem.
In the case of A1 0 -B C burnable poison rods, the linear heat generation rates aEe 3ve8y low in comparison to fuel rods (maximum local rate is less than 1.5 kW/ft).          '1here f ore , the stress analysis can be accomplished using conventional strength of                              0 3
materials formulae.                                                                        4 In the case of Gd 0 -UO      burnable poison rods, the peak poison rods will be operali$g al power levels lower than the peak power                            i to    rods, such that all analyses for fuel rods conservatively                            l bodnd that of the poison rods (Reference 41).                                              j The results of the burnable poison rod analyses confirm that the                            l design criteria of Section 4. 2.1.3.1, regarding stress, strain                            {
and strain fatigue, are satisfied.                                                      F j
4.2.3.3.2        Potential for Chemical Reaction A discussion of possible chemical reaction between poison material and the coolant was presented in Alo$ti8n            Se
                                                                                    -B C    B )
4.2.1.3.3.3, along with information on chemical compatibility between poison material and cladding.          Since the cladding material is    identical  to that of    the    fuel    rod    (Section 4.2.1.3.2),  the description of potential        chemical  reactions between cladding and coolant in Section 4.2.3.2.3 is applicable to both fuel and poison rods.
The potential for waterlogging rupture in A10 -B,C poison rods                          B is much lower than that in fuel rods becads2 oT the smaller                                6 thermal and dimensional changen that occur in a poison rod during reactor power increases.      Refer to Section 4.2.3.2.10 for a discussion of the potential for weterlogging rupture in fuel                              i rods.
4.2.3.4        control Element Asspably The CEAs are designed for a 15 effective full power year lifetime based on estimates for each CEA type of neutron absorber burnup,                        F allowable plastic strain of the Inconel 625 cladding and the resultant dimensional clearances of the elements within the fuel assembly guide tubes.
A. Internal Pressure l        For the Ag-In-Cd full-strength CEAs, no gas is released to the control rod void to contribute to internal g?s pressure. ,I Amendment F 4.2-68                  December 15, 1989
 
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0 For the twelve-element full strength CEAs, containing BC pellets, the value of internal pressure in the contr81 elements is dependent on the following parameters:
: 1. Initial fill gas pressure
: 2. Gas temperature
: 3. Helium generated and released
: 4. Available volume including B 4 C porosity of the absorber materials utilized in the CEA design, only the B C contributes to the total quantity of gas which must be ac,commodated within the control plemeng                        The polium ja produced by the nuclear reaction                      n      B
* Li  +3He ,
l and the fraction of the quantity gSner+                    a tid which is actcally 3
released to the plenum is temperature dependent and is predicted by the empirical equation discussed in Section 4.2.1.4.4.A.3.                    Temperatures used for release          fraction calculations are the maxinum predicted to occur during                                -
normal operation.                                                                      ]
The results of the CEA analyses confirm that the design criteria of Section 4.2.1.4, regarding stress, strain and 7
strain fatigue, are satisfied.
B. Thermal Stability of Absorber Materials None of'the materials selected for the control elements are susceptible to thermally induced phase changes at reactor                              "
operating conditions.                      Linear tl.ermal expansion, thermal conductivity, and melting points are given in Section i      4.2.1.4.
C. Irradiation Stability of Absorber Materials l      Irradiated        properties                of  the absorber materials are discussed in Section 4.2.1.4.                      Irradiation-induced chemical          ,
transmutations are produced in both the BC and the                                  F j
Ag-In-Cd. Neutron bombardment of B-10 atoms rekults in the L      production of lithium and helium.                        The percent of helium l
released is given by the expression in Section 4.2.1.4.
l Ag-In-cd alloy, which has an initial chemical composition of 79 wt% minimum Ag, 15 i 0.35 wt% In, 5 i wtt Cd and 0.2 wt%
maximum impurities, is expected to undergo small changes in composition.                Formation of 3 wt% tin due to the l
transmutation of indium and an increase in cadmium content to about 10 wtt due to the transmutation of silver is '
expected. These affect the thermal conductivity and linear expansion characteristics of the alloy and are accounted for in the design of the control elements.
Amendment F 4.2-69                December 15, 1989
_ _ . ~ ._. _ _ _ _ _ _ _ _ __
 
CESSAR tilhi..,                                                                                                      '
J Irradiation-enhanced                      swelling        characteristics                of    thel            l absorber    materials                    are      given      in            section      4.2.1.4. p        ;
Accommodations for swelling of the absorbers have been incorporated in the design of the control elements and                                                            ,
include the following measures:                                                                                    1
: 1. All B C pellets have rounded edges' to promote sliding                                                    ,
of thk pellets in the cladding due to differential                                                        ;
thermal expansion and irradiation enhanced swelling.                                                      ,
: 2. Dime.nsionally stable Type 304 stainless steel spacers are located at the bottom of all absorber stacks adjacent to the nose cap to minimize strain at the weld                                                    i joint.
: 3. A felt metal sleeve containing reduced diameter BC pellets is located in the bottom length of the absorb 8r stacks in full length control elements.                                        The felt metal sleeve laterally positions the reduced diameter BC pellets uniformly with respect to the clad and absor8s j
the  differential thermal expansion and irradiation                                                      ,
induced swelling of the B C pellets, thetreby limiting the. amount of induced stra$n the clad.
: 4. A hole is provided in the center of the Ag-In-Cd cylinder to accommodate swelling in excess of the                                                p amount expected over the life of the control element.
D. potential for and consequences of CEA Functional Failure The probability for a functional failure of the CEA is considered to be very small.                            This conclusion is based on the conservatism used in the design, the quality control procedures used during manufacturing and on testing of similar    full-size      CEA/CEDM                    combinations              under    simulated reactor conditions for lengths of travel and numbers of                                                          ,
trips greater than those expected to occur during the design life. The consequences of CEA/CEDM functional f ailure are discussed in Chapter 15.
A postulated CEA failure mode is cladding failure.                                          In the event that an elument is assumed to partially fill with j
water under low or zero power conditions, the possibility                                                        ,
exists that upon returning to power, the path of the water l      to the outside could be blocked.                                The expansion of the I      entrapped water could cause the element to swell.                                        In tests, specimens of CEA cladding were filled with a spacer l      representing the poison material.                                All but 9% of the remaining volume was filled with water. The sealed assembly was then subjected to a temperature of 650'F and an external Amendment F 4.2-70                              December 15, 1989
 
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l pressure of 2250 lb/in.2 followed by a rapid removal of the external pressure.                                  The resulting diametral increases of the cladding were on the order of 15 to 25 mils and were not sufficient to impair axial motion of the CEA, which has a 0.084 diametral clearance with the fuel assembly guide tubes.        This test result, coupled with the low probability of    a  cladding                                fmilure  leading  to  a  waterlogged rod, demonstrates that the probability                                      for a CEA  functional failure from this cause is low.
Another possible consequence of failed cladding is the release of small quantities of CEA filler materials, and helium and lithium (from the neutron-boron reactions).
However, the amounts which would be released are too small to have significant effects on coolant chemistry.
E. CEA Axial Growth Analysis Analysis has shown that adequate axial clearance exists between the bottom of the CEA finger and the fuel assembly t
guide tube.                                  This clearance, representative of the limiting    ,
design condition, has been calculated on. the basis of worst-case dimensional tolerances and considers the relative thermal growth between the fuel assembly and the fully inserted CEA.
4 4.2.4          TESTING AND INSPECTION PLAN Fuel    bundio assembly and control element assembly quality                                            ,_
assurance is attained by adherence to the procedures described in Chapter 17.
Vendor product certifications, process surveillance, inspections, tests,    and material check analyses are performed to ensure conformity of all fuel assembly and control element assembly components to the design requirements from material procurement through receiving inspection at the plant site. The following are basic quality assurance measures which are performed:
4.2.4.1                    Puel Assembly A comprehensive quality control plan is established to ensure that dimensional requirements of the drawings are met.                                    In those cases where a large number of measurements are required and 1004 inspection is impractical, the plan provides a high statistical confidence that dimensions are within tolerance. Sensitivity and accuracy of all measuring devices are within 110% of the dimensioned tolerance.                                    The basic quality assurance measures which are performed in addition to dimensional inspections and material verifications are described in the following sections.
i 4.2-71
 
CESSAR !!iWicm 4.2.4.1.1            Wald Quality Assurance Measures The welded joints used in the fuel assembly design are listed below in a series of paragraphs which describe the type and function of each veld and include a brief description of the testing (both destructive and non-destructive) performed to ensure the structural integrity of the joints.                                          The welds are listed from top to bottom in the fuel assembly.
The CEA guide tube joints (between the tube and threaded upper and lower ends)- are butt welds between the two Zircaloy subcomponents.          The welds are required to be full penetration welds and must not cause violation of dimensional or corrosion resistance standards.
The upper end fitting center guide post to lower cast flow plate joint has          a  threaded connection which                is              prevented      from unthreading by tack velding the center guide post to the bottom of the lower cast plate using the gas tungsten arc (GTA) process.
Each weld is inspected for compliance with a visual standard.
The spacer grid welds            at the intersection 'of perpendicular Zircaloy-4 grid strips are made by the GTA process.                                              Each intersection is welded top and bottom, and each weld is inspected by comparison with a visual standard.
For the spacer grid to CEA guide tube weld (both components                                                  ,
Zircaloy-4), each grid is welded to each guide tube with eight small welds, evenly divided between the upper and lower faces of the grid.          Each weld is required to be free of cracks and burnthrough and each veld is inspected by comparison to a visual standard.        Also, sufficient testing of sample welds is required
                  ,                          to. establish acceptable corrosion resistance of the weld region.
Each guide tube is inspected after welding to show that welding has not affected clearance for CEA motion.
The lower spacer grid welds at spacer strip intersections and between spacer and perimeter strips (all components Inconel 625) have the same configuration as for the Zircaloy spacer grids and                                          I are          all inspected  for  compliance            with    appropriate                    visual standards.
The lower spacer grid (Inconel) to Inconel skirt veld is made using the GTA process.              Each veld            is  inspected                      to ensure compliance with a visual standard.                                                                            ,
The Inconal skirt to lower end fitting (Type 304 stainless steel) weld is made using the GTA process and each weld is inspected to ensure compliance with a visual standard.
l' Amendment F 4.2-72                                        DeceEler 15, 1989
 
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l i
l The lower and fitting is fastened to the Zircaloy guide tubes                            j using threaded connections.
The connections are prevented from                    ,
unthreading by stainless steel locking discs which are welded to              'g the lower and fitting.        Each disc is tack welded to the and fitting in four places using the GTA process, and each weld is inspected for compliance with a visual standard.                                          i t
The inspection requirements and acceptance standards for each of                          i the welds are established on the basis of providing adequate                              i assurance  that    the  connections    will  perform  their    required functions.
4.2.4.1.2          other quality Assurance Measures All guide tubes are internally gaged,            ensuring free passage within the tubes, including the reduced diameter buffer region.                          ;
Each upper end fitting post to guide tube joint is inspected for                      s compliance with a visual standard.                                                    4 The spacer grid to fuel rod relationship is carefully examined at each grid location.                                ,
An alpha smear test is performed on the exterior surfaca of the
* fuel rods.
Each completed fuel assembly is inspected for cleanliness,                              -
wrapped to preserve its cleanliness and loaded in shipping                              -
t containers which are later purged and filled with dry air.
Visual inspection of the conveyance vehicle, shipping container, l    and fuel assembly are performed at the reactor site.            Approved                .
procedures are provided for unloading the fuel assemblies.
Following unloading, exterior portions of the fuel assembly components are inspected for shipping damage and cleanliness.            If
: j. damage is detected,        the assembly may be repaired onsite          or returned to the manufacturing facility for repair.        In the event                  '
the repair process is other than one normally used by the manufacturing facility, or that the repaired assembly does not meet the standard requirements for new fuel, the specific process or assembly is reviewed before it is accepted.        Each spacer grid is checked to verify compliance with outside dimension, grid cell pitch, and spring tab preset requirements.
4.2.4.2        Fuel Rod 4.2.4.2.1          Fuel Pellets l    During the conversion of UF          to ceramic grade uranium dioxide powder,  the UO 2    Powder is  ' divided into  lots blended to form Amendment B                        l 4.2-73                March 31, 1988 1
1
 
CESSARIMe.n.
t uniform isotopic, chemical, and physical characteristics.                                                                    Two containers are selected from the total number of containers in                                                                  '
each lot for certification sampling. Samples are removed from each of the two selected containers and subdivided to verify specification limits (section 4.2.1.2.4).
Pellets are divided into lots during fabrication with all pellets within the lot being processed under the same conditions.
Representative samples are obtained from each lot for product acceptance tests.                      Total hydrogen content of finished ground                                                l pellets is restricted (Section 4.2.1.2.4.1).
The        pellet    diameters            are                100%        inspected;  all    other pellet                      [
dimensions meet a 90/90 confidence lvel.                                            Density requirements of the sintered pellet (Section 4.2.1.2.4.3) must meet a 95/95 confidence level.                    Longitudinal sections of two sample pellets from each pellet lot are prepared for metallographic examination to ensure conformance to microstructure requirements (section 4.2.1.2.4.2). Pellet surfaces are inspected for chips, cracks, and fissures in accordance with approved standards.                                                                              '
4.2.4.2.2                      Cladding Lots are          formed            from tubing produced                          from the    same                    ingot, annealed in the same final vacuum annealing charge and fabricated using the same procedures.                                    Samples randomly selected from each lot of          finished tubing are chemically analyzed to ensure conformance to specified chemical requirements, and to verify tensile properties and hydride orin tation.                                            Samples from each lot are also used for metallographic and burst tests.                                                                      Each finished tube is u1~'rasonically tested over its entire length for internal soundness; visually inspected for cleanliness and the absence of acid scains, surface defects, and deformation; and, inspected for inside dimension and wall thickness. The following summarizes the test requirements:
A.        Test (refer to Section 4.2.1.2.2) l                  1. Chemical Analysis I
Ingot analysis is required for top, middle, and bottom of each ingot.                Finished product is tested for hydrogen, nitrogen, chrbon, and oxygen per ASTM E146-68.
I                  2    Tensile Test at Room Temperature (ASTM EB-69)
: 3. Corrosion Resistance Test (ASTM G2-67)
: 4. Grain Size (ASTM E112-63) l 4.2-74 l
l
 
C E S S A R tl W ie m ..                                                                                                  !
: 5.            Hydrostatic Burst Test (section 4.2.1.2.2.1) 6              Surface Roughness
: 7.            Visual Examination                                                                                    i
: 8.            Ultrasonic Test
: 9.            Wall Thickness                                                                                        )
: 10.            Straightness                                                                                          '
: 11.            Inside Diameter 4.2.4.2.3                        Fuel Rod Assembly Immediately prior to loading, pellets must be capable of passing approved visual standards.                      Each fuel pullet stack is weighed to within 0.1%                    accuracy. The loading process is such that                                          x cleanliness and dryness of all internal fuel rod components are                                                        e maintained until after the final and cap weld is completed.                                                              e Loading and handling of pellets is carefully controlled to minimize chipping of pellets.
The following procedures are used during fabrication to assure that there are no axial gaps in fuel rodst 4.2.4.2.3.1                          stack Length Gage                                                                    ,
The operator stacks pellets onto V troughs that are gage marked                                                        -
to the proper. fuel column height.                          When pellet stacking                                  is completed, all column heights are checked by Quality Control.
l    The pellets are subsequently loaded into tubes. After loading, j
the distance from the end -of the tube to the end of the pellet
:    column is checked with a gage.
i 4.2.4.2.3.2                          Fluoroscopy Finished fuel rods, prior to being loaded into assemblies, are fluoroscoped to ensure that no significant gaps exist in the fuel Column.
Loaded fuel rods are evacuated and backfilled with helium to a prescribed level as determined for the fuel batch.                                                        Impurity content of the fill gas shall not exceed 0.5%.
The fuel rod and cap-to-fuel rod cladding tube welds will be butt welds between the Zircaloy-4 cladding tube and the Zircaloy-4 end l    cap machined from bar stock. The weld process will be magnetic force welding (HFW) .                    Quality assurance on the end cap weld will include:
4.2-75
 
1 C E S S A R II M ,cu .                                                                                                    ;
i A. Destructive examination of a sufficient number of weld samples to establish that the maximum allowable percent of unbonded wall thickness (15%) and the maximum allowable continuous unbonded region (10%) are not exceeded.
                                                                                                                                ]
B. Visual examination of all and cap welds to establish freedom from cracks, seams, inclusions and foreign particles after                                                          !
final machining of the weld region.                                                                                  ;
l C.-                                                                p welds to establish that                              I Helium no leak leak rate checking greater than of all          10 ,ydcm  cp/s is present.
D.-  Corrosion testing of a sufficient number of samples to                                                              )
establish that weld zones do not exhibit excessive corrosion                                                          '
compared to a visual standard.                              Welds must be capable of passing a corrosion test (ASTM G2) with no prefereptial                                                            ,
oxidation at the weld in water at 650*F, 2200 lb/in,                                                  for        ;
                                                                                                                              ~
3-1/2 days.
\
!    All finished fuel rods are visually inspected to ensure a proper surface finish (scratches greater than 0.001 inch in depth,                                                              !
cracks, slivers, and other similar defects are not acceptable).
Each fuel rod is marked to provide a means of identification.
4.2.4.3          nurnable Poison Rod 4.2.4.3.1          Burnable Poison Pellets For the fabrication of A10 3-B C pellets, B C powder is sampled                                                          f to verify particle size a$d wtk boron requifements prior to its use in pellet production.                Finished pellets are 100% inspected for diameter and must satisfy a 90/90 confidence level on other dimensions. Samples are taken from each of the pellet lots and examined for uniform dispersion of the B C in Al O                                                  Conformance with density range requirements is dmonstrakeh. at a 95/95 confidence level and with BC                  4 requirements at a 90/90 level.                                .
  . Samples are drawn from each 1ot to verify acceptable impurity levels. Finally, all pellets are inspected for conformance with surface chip and crack standards.
The fabrication of Gd 0 -UO pellets is essentially the same as for UO f and U$ uel    pelletssize.
particle      e$cipt korRestrictions  tighter restrictions          on the GdonOkhd are introduced                              B particl$ size to promote homogeneity of the Gd 0 -UO mixture.
For this poison application, natural UO. wilf 3)e dsed as a carrier for the gadolinium. The fabricati@, of Gd 0 -UO pellets employs dry blending and mixing of the necessar/ handities of UO 2  and Gd 23 0    powders. As              with UO2 pellets, these powders are Amendment B 4.2-76                                  March 31, 1988
 
CESSARIBLn.                                                                                                                            )
f i
then pelletized by blanding and sintering processes similar to                                                                        i those employed in the manufacture of UO pellets. The sintering process promotes formation of a solid s$1ution of Uo and Gd o .
As with UO must meet,2 pellets, the Gd o -UOstringent                                        on density,    specifit:alion$
grain size,    pellets and Bare tesEed homogeneity.      In particular,                                              the density and densification                            l specifications (4 T.D.), grain size requirements and blending                                                                          i requirements are essentially the same as for a Uo mixture.                                                                            ;
2                                ,
4.2.4.3.2          Cladding The testing and inspection plan for burnable poison rod cladding is identical to that for fuel rod cladding (Section 4.2.4.2.2).
The moisture content of poison pellets prior to loading                                                                    is limited to values below that which would be required to produce primary hydride penetration of the cladding.                                                          Total moisture inventory is comparable to that which has been shown to be acceptable in fuel rods (Reference 60).                                                      The fabrication process                  ,
is such that all steps from component drying through final                                                                          2 l
welding are      carefully controlled                                              so as to minimize                    the possibilities for excessive moisture pickup.                                                            For Al,0,-B,C poison rods, final verification of pellet dryness is nthea By destructive examination of one poison rod for each group of rods                                                                        l from the same drying lot.                                                  For Gd,o -Uo, poison rods, final verification of pellet dryness is fddntidal to that for fuel                                                                  B rods.                                                                                                                                  !
The following procedure is used during fabrication to assure that there are no axial gaps in poison rods,                                                                                                  l The operator stacks pellets onto V troughs that are gage marked to the proper column height. When pellet stacking is completed, all column heights are checked by Quality control. The pellets are subsequently loaded into tubes.                                              After loading, the distance                            ,
from the end of the tube to the end of the pellet column is                                                                              1 checked with a gage.                                                                                                                    )
Loaded poison rods are evacuated and backfilled with helium to a prescribed level.                    Impurity content of the fill gas must not exceed 0.5%.
End cap weld integrity and corrosion resistance is ensured by a Quality Control plan identical to that used in fuel rod fabrication (Section 4.2.4.2.3).
Each poison rod is marked to provide a means of identification.
Amendment B                        l 4.2-77                      March 31, 1988
 
CESSAR MLano 4.2.4.4                        C.ontrol Element Assamhlies The CEAs are subjected to numerous inspections and tests during manufacturing and after installation in the reactor. A general product specification controls the fabrication,                                                inspection,      i assembly, cleaning, packaging, and shipping of CEAs.                                                    All materials are procured to AMS, ASTM or C-E specifications.                                              In addition, various CEA hardware tests have been conducted or are in progress.                                                                                                    ,
During manufacturing, the following inspections and tests are                                                    ,
performed:                                                                                                      ,
A.          The loading of each control element is carefully controlled to obtain the proper amounts and types of filler materials for each type of CEA application (e.g., full-strength B4 C or                                p Ag-In-Cd).
B.-          All end cap welds are liquid penetrant examined and helium leak tested. A sampling plan Ls used to section and examine and cap welds.
C.          Each type of control element has unique external features which distinguish it from other types.
D.          Each CEA is serialized to distinguish it from the others.
See Figures 4.2-3 through 4.2-5.
E.          Fully essembled CEAs are checked for proper alignment of the neutron absorber elements using a special                                      fixture. The alignment check ensures that the frictional force that could result from adverse tolerances is below the force which could significantly increase scram time.                                                      B In        addition              to  the        basic measures discussed above, the manufacturing process includes numerous other quality control steps for ensuring that the individual CEA components satisfy design requirements for material quality, detail dimensions, and process control.
After installation in the reactor, but prior to criticality, each CEA is traversed through its full stroke and tripped. A similar procedure will also be conducted at refueling intervals.
l The required 90% insertion scram time for CEAs is 4.0 seconds under worst case conditions.                                Verification of adequacy was                  p initially determined by testing in the C-E TF-2 flow test facility as reported in Appendix 4B.                                            This test facility contained              prototypical                  (System          80)    reactor          components consisting of fuel assemblies, CEA shroud, control element drive l
Amendment F 4.2-78                            December 15 1989
 
CESSARtm%am.                                                                                                                  I i
i l
mechanism, and a simulation of surrounding core internal support                                                              I components.      The    test        conditions          simulated the range of                                              '
temperatures and flow rates predicted for System 80 normal plant operation. The required scram time has been subsequently verified to be conservative by testing at the Palo Verde (System F
: 80) operating units.
4.2.5        REACTOR INTERFACE REQUIREMENTS (The  interface requirements are retained from CESSAR-F for                                                                I continuity. In the completed CESSAR-DC they will be superseded.
Below are detailed the interface requirements that the reactor places on certain aspects of the BOP, listed by categories. In addition, applicable GDC and Regulatory Guides which C-E utilizes in its    design    of  the        reactor are presented.                        The GDC and Regulatory Guides are listed only to show what C-E considers to be relevant, and are not imposed as interface requirements unless specifically called out as such in a particular interface requirement.                                                                                                              e t
Relevant GDC - 1, 2,            3,  4, 10, 11, 12, 14, 15, 26, 27, 28, 29, L
30, 31, 32, 61, 62, 63 Relevant      - 1.2, 1.13, 1.20, 1.25, 1.28,                      1.29,            1.31,    1.34, Reg. Guides      1.36, 1.37, 1.38, 1.43, 1.44,                    1.50,            1.54,    1.60,    B 1.61, 1.64, 1.65, 1.68, 1.71, 1.74, 1.84, 1.85                                                          g .
A. Protection from Natural Phenomena w
: 1. High    winds, tornado, tornado missile and flooding requirements relating to the reactor are in accordance with Criterion 2 of 10 CFR 50 Appendix A.
: 2. The spent fuel pool shall be a seismic Category I structure.
: 3. The load bearing members of the spent fuel storage racks shall withstand the forces induced by the SSE vertical and horizontal seismic loadings. These forces shall    be          assumed    as          acting  simultaneously                  in conjunction with the combined deadweight and live loads without exceeding minimum material yield stresses as specified by ASME Section III Subsection NF.                                                B
: 4. The    spent          fuel  storage          racks  shall              be    seismic Category I.
Amendment F 4.2-79                              December 15 1989
 
CESSAR timfiew.
B.                                Protection from Pipe Failure
: 1. The fuel shall be protected from the effects of pipe whip while in storage.
: 2. Refer    to  Section 5.1.4 for            protective  measure requirements for the reactor.
: 3. Spent fuel shall be protected from the effects of pipe rupture.
: c.                                Missiles
: 1. A  removable structure shall be located above the reactor vessel to block any missile that could be generated by a control element drive mechanism.
: 2. The fuel shall be protected        from. the        effects  of missiles while in storage.
D.                                Separation
: 1. New Fuel Storage Racks
: a. The new fuel storage racks shall be designed such that fuel assemblies will not be inserted in other than prescribed locations.
: b. Adequate margin to criticality shall be provided for full rack loadings of fuel assemblies having a mechanical design similar to that described in this Chapter and enrichments up to 5.0 w/o U-235.          F
: c. The degree of suberiticality provided shall be consistent with the requirements of ANSI Standard N18.2 (Section 5.7.4.1).
E.                                Thermal Limitations
: 1. Cooling air shall be provided to the CEDMs at a minimum flow rate of 700 SCFM per CEDM at a temperature in the range of 80-120*F.
: 2. Drains,    permanently  connected systems, and other features of the spent fuel pool shall be designed so that neither maloperation nor failure can result in loss of coolant that would uncover the stored fuel.
Amendment F 4.2-80                      December 15 1989
  - _ _ _ _ - _ _ _ _ - . _ - _ - _ - . . . - , _ _ ~                            .  -  . - . .
 
CESSAR tilanm..                                                                                                                          ;
: 3. Spent fuel pool cooling capacity shall be consistent with the design basis in Section 9.1.3.1.                                                                              F F. Monitoring
: 1. Low water level alarms shall be provided for- the                                                                            I refueling pool and the spent fuel pool.                                                                                    J
: 2. A system shall be provided to monitor the Reactor Coolant System for internal loose parts.                                                  The system                        )
shall have the ability to detect a loose part striking                                                                      )
the              internal      surface              of  Reactor Coolant                      System                        i components with an energy level of one-half foot pound or more.                  The system shall have alarm and recording                                                        !
capability.                  The system design shall be suitable for the temperature and humidity environment experienced in the area where the equipment normally operates.
G. Inspection and Testing
: 1. In-service                  Inspection              (ISI)  shall            be        performed      in accordance with Section XI of the ASME Code.
* H. Materials
: 1. See Section 5.1.4.L.3.                                                                                                  .
I. Related Services                                                                                                                  ;
: 1. For refueling operations, the containment building crane shall have a minimum capacity of 225 tons.
: a.            A hoisting speed of 6 inches per minute or less i                                          shall be utilized during-refueling operations.
: b.              A load measuring device shall be provided for use during heavy lifts.
: c.            A low inching speed                      is required during                    those portions of the lift when close tolerance surfaces are engaging each other.
: 2. An overhead crane shall be provided in the new fuel storage area to facilitate handling of new fuel,
: a.            The crane capeity shall be at least 1 ton to accommodate the weight of a fuel assembly.
l l                                                                                                                Amendment F 4.2-81                                            December 15 1989
 
CESSARtmhr.
: b. A vertical hoisting speed of 6 feet / minute or less shall be provided.
: c. The crane load shall be capable of being limited to prevent the hoist load from exceeding 5000 pounds when handling fuel assemblies.
: 3. See Section 5.1.4.P.3.
I 4.2-82
                                                    . . _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ . . _ _ _ _ _ _a
 
CESSAR RHnnemo l
I pturzazucas roa sacTrom 4.a
'~
: 1. Timoshenko,        S., Strancrth of Materials, Part II Chapter IX, D. VanNostrand Co., Inc., New York, 1956.
: 2.  "High Temperature Properties of Zircaloy and UO                              for use in IOCA        Evaluation Models,"                Combustion Enginekring, Inc, CENPD-136 (Proprietary).                                                                      ;
: 3.  "Zircaloy          Growth-In-Reactor                Dimensional            Changes  in    j Zircaloy-4 Fuel Assemblies," Combustion Engineering, Inc.,                                    !
CENPD-198P (Proprietary), December 1975.
: 4.  " Extended Burnup Operation of Combustion Engineering PWR
;      Fuel,"            Combustion        Engineering,            Inc.,          CENPD-269-P B l        (Proprietary), July 1984.
: 5. O'Donnel, W. J., " Fracture of Cylindrical Fuel Rod cladding due to Plastic Instability", WAPD-TM-651, April 1967.
: 6. Weber, J. M. " Plastic Stability of Zircaloy-2 Fuel Cladding, Effects on Radiation of Structural Metals," ASTM STP 426, Am. Soc. Testing Mats., pp 653-669, 1967.
: 7. Engle, J. T.          and Meieran, M. B.,            " Performance of Fuel Rods              i Having 97 Percent Theoretical Density UO, Pellets Sheathed r        in  Ziroaloy-4        and  Irradiated            at  Low " Thermal        Ratings,"
l
      . WAPD-TM-631, July 1968.
: 8. Duncombe, E., Meyer, J. E., and Coffman, W. A., " Comparisons with Experiment of Calculated Dimensional Changes and Failure Analysis of Irradiated Bulk Oxide Fuel Test Rods                                    ,
Using          the    CYGRO-1        Computer          Program,"          WAPD-TM-583, September 1966.
: 9. McCauley, J. E.,            et al.,          " Evaluation    of    the      Irradiation Performance of Zircaloy-4 Clad Test Rod Containing Annular UO 2
Fuel Pellets (Rod 79-19)," WAPD-TM-595, December 1966.                                  ;
: 10. Notly, M. J. F., Bain, A. S., and Robertson, J. A. L., "The Longitudinal and Diametral Expansion of UO Fuel Elements,"      2 AECL-2143, November 1964.
: 11. Notley,        M. J. F.,  "The Thermal Conductivity of Columnar Grains          in  Irradiated            U0 2
Fuel    Elements,"          AECL-1822, July 1962.
1 Amendment B 4.2-83                              March 31, 1988
 
C E S S A R W % m ..
i l
i
: 12. Hanson, S. S.,    " Fatigue    A Complex Subject                -
Some Simple              l Approximations," Ernerimental Mechanics, Vol. 22, No.                        2,  pp          {
193-226, July 1965.
: 13. O'Donnel, W. J. and Langer, B. F., " Fatigue Design Basis for Zircaloy Components," Nuc. sei. Ena., Vol 20, pp 1-12, 1964.                                    1
: 14. CESSAR Proprietary Appendix, Docket 50-470.
: 15. "C-E  Fuel  Evaluation Model      Topical                Report,"    Combustion Engineering,    Inc.,    CENPD-139-P    (Proprietary),                CENPD-139 Rev. 01 (Non-Proprietary), CENPD-139 Supplement                      1,  Rev. 01 (Non-Proprietary), July 1974.
: 16. " Improvements    to    Puel    Evaluation Model," Combustion                                ,
Engineering, Inc., CEN-161-P(A) (Proprietary), August 1989;
:    and CEN-161-P(B) Supplement 1 (Proprietary), April 1986.                                      ;
l                                                                                            F
: 17. " Fuel  Rod  Maximum      Allowable    Gas      Pressure,"          combustion Engineering, Inc., CEN-132-P (Proprietary), June 1988.
: 18. Conway, J. B., "The Thermal Expansion and Heat Capacity of                          '
UO p to 2200'C," CE-NMPD-TM-63-6-6.                                                        +
: 19. Christensen,  J. A.,  " Thermal Expansion of UO 2," HW-75148, 1962.
: 20. Jones, J. M.,      et al.,    " Optical      Properties            of    Uranium Oxides," Nature, 205, 663-65, 1965.
: 21. Cabannes, F. and Stora, J. P., " Reflection and Emission Factors of UO., at High Temperatures," C. R. Acad. Sci.,
Paris, Ser. B.264 (1) 45-48, 1967.
: 22. Held, P. C.      and      Wilder, D. R.,                  "High    Temperature Hemispherical Spectral Emittance of Uranium Dioxide at 0.65 and 0.70m," J. Am. Cer. Soc., Vol 52, No.                    4,  1969.
l l 23. Bransfield, M. C.,      " Recommended      Property            and      Reaction Kinetics Data for Use in Evaluating a Light Water Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy-4 or 324-53 Clad Uo ," GEMP-482, 1968.
2
: 24. Beals, R. J., "High Temperature Mechanical Properties of                                        I Oxide Fuels," ANL-7577, April-May 1969, Page 160.
l                                                                                                    l
: 25. "CEPAN,  Method of Analyzing            Creep              Collapse      of    Oval l    Cladding," Combustion Engineering,                Inc.,        CENPD-187      P-A,            l March 1976.                                                                                    l 1
l i                                                                                                    l I                                                                      Amendment F                    !
December 15, 1989              l 4.2-84 l
                                          ..    -    - . - - .-                              _0
 
CESSAR tilh
: 26.  "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program,"      Combustion      Engineering,      Inc.,    CENPD-13SP (Proprietary), CENPD-135 (Non-Proprietary), August 1974.
: 27. Deverall, J. E., IA-2669, USAEC, Vol 62, 1954.
: 28. Rudkin, R. L.,          Parker, J. W.,      and      Jenkins, R. J.,
ASD-TDR-62-24, Vol 1, pp 20, 1963.
: 29. Thorne, R. P. and Howard, V. C., " Changes in Polycrystalline Alumina by Fast Neutron Irradiation," p 415, Proceedinas of the British Ceramie Society, No. 7, February 1967.
: 30. Sinnad,  M. T. and Meyer, R.      S., " Boo Review of Properties for Nuclear Reactor          Applications,"    Proceedinas of the conference    on    Nuclear    Aeolications    of    Nonfissionable Ceramics, pp 209-210, May 9-11, 1966.
: 31. Rason, N. S. and Smith, A. W., NAA-SR-862, Vol 37 (AD85006),            a 1954.
: 32. Saba, W. G. and Sterret, K. F.,        J. Am. Chem. Soc., Vol 79, pp 3637-38.
: 33.  " Fuels and Materials Development Quarterly Progress Report,"
pp 38-58, ORNL-TM-3703, December 31, 1971.
: 34. Kingary, W. D., Introduction to ceramics, John Wiley T. Sons, pp 486-504.
: 35. Toulookan,    Y. S. Thermochvsical    Pronerties    of  Hiah Tameerature Solid Materials; Vol 4 and 5, MacMillan.
: 36. Moore,  G. E. and Kelley, K. K., J. Am. Chem. Soc., Vol 69, pp 309-16, 1947.
: 37. Keilholtz,    G. W.,  Moore,  R. E.,    and Robitson,    M. E.,
        " Effects of High Boron Burnups on B C and ZrB, Dispersion in Al 0 and Zircaloy-2," BM1-1627, Aphil 24, 1963.
23
: 38. Burian,  R. J., Fromm, E. O., and Gates, J. E. "Effect of High Boron Burnups on B C and ZrB, Dispersions in Al 23        0 and Zircaloy-2" BM1-1627, A ril 24, 1W3.
: 39. Cunningham, G. W.,      " Compatibility of Metals and Ceramics,"
Proceedinas    of    Nuclear  Aeolications    of    Nonfissionable Ceramics, pp 279-289, May 1966.
4.2-85
 
CESSAR timneau.=
: 40. Graber, M. J., "A Metallurgical Evaluation of Simulated BWR Emergency Core Cooling Tests," Idaho Nuclear Corporation, IN-1453, March 1971.
: 41.  "C-E Methodology for Core          Designs  Containing Gadolinia Urania Burnable Absorbers,"        Combustion Engineering, Inc.,    B CENPD-275-P (Proprietary), March 1987.
: 42. Pitner, A, L.,    "The WDC 1      Instrumental Irradiation of Boron    Carbide      in    a    Spectrum-Hardened    ETR  Flux,"
MEDL-TME-73-38, April 1973.                                          p r
: 43. Gray, R. G. and Lynam, L. R. , " Irradiation Behavior of Bulk BC    and    B C-Sic Burnable Poison Plates,"              WAPD-261, O tober 1963
: 44.  "HEDL Quarterly Technical Report for October, Nover.ber and December 1974,"      Vol 1, HEDL-TME-74-4, pp A-51 to          A-53, January 1975.
: 45. Mahagan, D. E.,      " Boron    Carbide    Thermal'  Conductivity,"
HEDL-TME-73-78, September 1973.
: 46. Homan, F. J., " Performance Modeling of Neutron Absorbers,"
Nuclear Technoloav, Vol 16, pp 216-225, October 1972.
: 47. Pitner, A. L. and Russcher, G. E., " Irradiation        of Boron    i Carbide Pellets and Powders in Hanford Thermal Reactors,"
WHAN-TR-24, December 1970.
: 48. Pitner, A. L. and Russcher,    G. E.,  "A Function on Predict LMFBR Helium Release Bound on Boron Carbide Irradiation Data from Thermal Reactors," HEL-TME-71-127, September 30, 1971.
: 49.  " Materials Technology Program Report for October, November, and December 1973," HEDL-73-6, pp A-69 to A-72.
: 50. Cohen, I., "Devalopment and Propertiss of Silver-Base Alloys as Control Rod Materials for Pressurized Water Reactors,"
WAPD-214, December 1959.                                              F
:i
: 51. Tipton,    C. R.,    " Reactor Handbook,"    Vol. 1,  Materials, Interscience, p. 027, 1960.
4 Amendment F 4.2-86                Decenter 15 1989
 
1 CESSAR !!Nne.no
                                                                                                                                )
I l
: 52. " National Alloy Development Program Information Meeting," pp                                                        ;
39-63, TC-291, May 22, 1975.
                                                                                                                                ]
: 53. " Quarterly        Progress                    Report  -
Irradiation                    Effects    on            )
Structural Materials," HEDL-TEM-161, pp GE GE-10.                                                                  l
: 54. " Fuel Performance Evaluation in 16x16 Assemblies at Arkansas Nuclear One, Unit 2, Final Report," EPRI NP-4250 M, Project                                                          I 586-1, March 1986.
B
: 55. The Evaluation and Demonstration of Methods for Improved                                                          p    i Nuclear    Fuel                  Utilization,          Eleventh                    Progress    Reportt January 1, 1987 to September 30, 1988, DOE /ET/34013-14, July                                                        l 1989.                                                                                                          l    ;
: 56. " Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading," Combustion Engineering, Inc.,
CENPD-178, Rev. 1, August 1981.
: 57. " Fuel and Poison Rod Bowing," Combustion Engineering, Inc.,
CENPD-225-P (Proprietary), October 1976.                                                                          p
: 58. " Application of Zirceloy Irradiation Growth Correlations for the Calculation of Fuel Assembly and Fuel Rod Growth Allowances," Supplement 1 to CENPD-198-P,                                                (Proprietary),
L          December 1977.
: 59. Pickman, D. O., " Properties of Zircaloy Cladding," Nuclear                                                        [
Enaineerina and Desian, Vol. 21, No. 2 (1972).
l
: 60. Joon, K., " Primary Hydride Failure of Zircaloy Clad Fuel Rods," ANS Transactions, Vol 15, No. 1.                                                                              ,
: 61. Caye, T. E.          "Saxton Plutonium Project, Quarterly Progress Report for the Period Ending March 31, 1972," WCAP-3385-31,
* November 1972.
: 62. Berman, R. M.,                        Meieran, H.        B.,              and          Patterson, P.,
          " Irradiation Behavior of Zircaloy-Clad Fuel Rods Containing l-        Dished End UO Pellets," (LWBR-LSBR Development Program),
WAPD-TM-629, Jbly 1967.
: 63. Baroch, S. J.,              et al.,            " Comparative Performance of Zircaloy l          and Stainless                  Steel Clad            Fuel Rods Operated to 10,000 l          mwd /MTU in the VBWR," GEAP-4849, April 1966.
l l
: 64. Megerth,  F.          H.,              "Zircaloy-Clad UO 2              Fuel        Rod  Evaluation l          Program,"            Quarterly                    Progress                  Report          No.      8, August 1969-October 1969, GEAP-10121, November 1969.
Amendment F 4.2-87                                  December 15 1989
 
i CESSARtm%m.
I
: 65. Megerth, F. H., "Zirealoy-Clad UO                                                      Tuel Rod Evaluation Program,"                      Quarterly          Progress 2                          Report      No.              1,    ,
                                                                                                                                      )
November 1967-January 1968, GEAP-5598, March 1968.                                                                            i
: 66. Garde, A. M., and Pati, S.                                R.,                  " Gas Release Densification, Swelling and Microstructural Evaluation of Four-cycle Tual                                                                  l Rods from Calvert Cliffs-1," C-E NPSD-211, April 1983.                                                                      i t
: 67. San Onofre Nuclear Generating Station, Units 2 &                                                      3,  Final Safety Analysis Renort, Volume-9, pages 4.2 4.2-61.
: 68. Brite, D. W. et al, "EEI/EPRI Fuel Densification Program Final Report," Battelle Pacific Northwest Laboratories, March 1975.
: 69. J. C.            LaVake and M. Gartner, "High Burnup PWR Ramp Test                                                      0 Program, Final Report," DOE /ET/34030-10, December 1984.
: 70. Andrews, M. G., Smith, G. S., and Garde, A. M.,                                                    " Experience and Developments with Combustion Engineering                                                      Fuel," ANS Topical Meeting on LWR                            Tuel                          Performance,  Williamsburg,                !
Virginia, April 1988.
: 71. Stephan, L.                    A., "The Response of Waterlogged UO 2
Fuel Rods to Power Bursts," IDO-ITR-105, April 1969.
: 72. Stephan, L. A., "The Effects of Cladding Material and Heat Treatment on the Response of Waterlogged UO2 Fuel Rods to Power Burst," IM-ITR-111, January 1970.
: 73. " TORC Code:                    A Computer Code for Determining the Thermal                                                !
Margin of a Reactor Core," Combustion Engineering, Inc.,
CENPD-161-P, (Proprietary) July 1, 1975.
i Amendment B 4.2-88                                          March 31, 1988
 
e                                                                                                                                                              i CESSAR WWieu Intu.1                                                                                          )
(Sheet 1of6)                                                                                          l 1
BEH&KKALE116tLP.ARAMEIB1                                                                                    i I
Core Arraneament Number of fuel assemblies in core, total                                                                    241 Number of CEAs                                                                                              93                              B        l Number of fuel rod locations                                        .
56,876 Spacing between fuel assemblies, fuel rod surface to                                                                                                ,
surface inches                                                                                          0.208                                  :
Spacing, outer fuel rod surface to core shroud, inches                                                      0.214                                  '
Hydraulic diameter, nominal channel, feet 2                                                                  0.0393 Total flow area ( giudingguidetubes),ft                                                                      60.9                                  i Total core area,                                                                                            112.3                                  :
Core equivalent diameter, inches                                                                            143.6                            s    2 Core circumscribed diameter, inches                                                                          152.46                            :    <
Total fuel loading, kg U (assuming all rod locations are fuel rods)                                                                                              102.7 x 10 3 Total fuel weight, Ib U02 (assuming all rod locations are fuel rods)                                                                                              257.1 x 10 3 Total weight of Zircaloy,1b                                                                                  71,758 l            Fuel volume (including dishes), ft 3                                                                        409.6 Fuel Assemblies e,
No. of                                Enrichment                          No. of Poison Rods IAlsh          Assemblies                                  (wt0 U-235                            ner Assembly A0                  69                                      1.92                                            0 B1                  44                        12 rods with 1.92                                            16 t
208 rods with 2.78 B2                  64                        12 rods with 1.92                                            16 208 rods with 2.78 CO                  40                        12 rods with 2.78                                              0 224 rods with 3.30 C1                  24                        12 rods with 2.78                                            16 208 rods with 3,30 l
Fuel Rod Array                                                                        square, 16 x 16 Fuel Rod Pitch, inches                                                                0.506 Amendment B March 31, 1988
 
l C E S S A R t mir.cui.
TABLE 4.21(Cont'd)
(Sheet 2of6)
MECHANICAL DESIGN PARAMETERS Fuel Assemblies (Cont'd)
Spacer Grid ic  Type                                                                  Leaf spring Material                                                              Zircaloy-4 Number per assembly                                                    10 Weight each, Ib                                                        1.8 I
Bottom Spacer Grid Type                                                                  Leaf spring                                '
Material                                                              Inconel 625 Number per assembly                                                  1 Weight each, lb (with skirt)                                          3.2 Weight of fuel assembly,1b                                                  1436 Outside dimensions Fuel rod to fuel rod, inches                                          7.972 x 7.972 Fuel Rod                                                                        .
l Fuel rod material (sintered pellet)                                  UO Pellet diameter, inches                                              0$25 Pelletlength,incheg                                                  0.390 i          Pellet density, g/cm                                                  10.38 o          Pellet theoretical density, g/cm 3                                    10.96
!          Pellet density (% theoretigal)                                        94.75 l          Stack height density, g/cm                                            10.061 Clad material                                                          Zircaloy-4 Clad ID, inches                                                        0.332 Clad OD, (nominal), inches                                            0.382 l'        Clad thickness, (nominal), inches                                      0.025 Diametral gap, (cold, nominal), inches                                0.007 Active length, inches                                                  150 Plenum length, inches                                                  7.918                                7 Amendment F December 15, 1989
 
1 CESSAR timr.cui.
TABLE 4.2-1(Cont'd)
(Sheet 3 of 6)
NECHANICAL DESIGN PARAhETERS Control Element Assemblies (CEAs)
Twelve-Element                                      Four-Element                                ;
Full Strenoth CEA                                                                                                                                        B Number                                                      48                                                  20 Absorber elements, No. per assy.                            12                                                  4 Type                                                        Cylindrical                                        Cylindrical rods                                                rods Clad material                                              Inconel 625                                          Inconel 625 Clad thickness, inches                                      0.035                                              0.035 Clad 00, inches                                            0.816                                              0.816                                        r Diametral gap, inches                                      0.009                                              0.012 Elements                                                                                                                                                  F Poison material                                          B C/ Felt metal                                    Ag In Cd                                    '
a$dreduceddia.
BC 4
Poison length, inches                                    135.5/12.5                                          148 B4 C Pellet Diameter, inches                                          0.737/0.674                                          N/A                                      B Density, % o{ theoretical density                        73                                                  N/A of 2.52 g/cm Weight % boron, minimum                                    77.5                                                N/A Ag In Cd Cylindrical Bar with Central Hole F
Outside diameter, inches                                  N/A                                                  0.734 Inside diameter, inches                                  N/A                                                  0.25 l    Length of bar, inches                                    N/A                                                  2 l
Amendment F December 15, 1989
 
i l
CESSARti h n..                                                                                                          i l
1 TABLE 4.2-1(Cont'd)                                                                      l 1
($heet4of6)
MECHANICAL DE$ltN PARAMETERS Control Element Assemblies (CEAs) (Cont'd)
Four-Element                                    i i
Part Strenath CU                                                                                                      l Number                                                                25 Absorber elements, No. per assy.                                      4 Type                                                                  Cylindrical rods Clad material                                                        Inconel 625 F
Clad thickness, inches                                                0.035 Clad 00, inches                                                      0.816                                            I Diametral gap, inches                                                0.009 Elements                                                                                                              .
Poison material                                                    Inconel 625 Poison length, inches                                              148
  -Inconel 625 Cylindrical Bar Diameter, inches                                                    0.737 Length of bar, inches                                              2 Burnable Poison Rod Alumina - Boron Carbide Poison Rod Desian Absorber material                                                    A123 0 -B4C                                    ;
Pellet diameter, inches                                              0.307 Pellet length, inches, min                                            0.500 Amendment F December 15, 1989
 
                                                                                                                                                          ^
1 I
CESSAR timf.co...                                                                                                                            !
I l
TABLE 4.21(Cont'd)                                                                                      l (Sheet 5of6)                                                                                      j K Gt M N                                                                                                    ,
turnable Poison Rod (Cont'd)-                                                          _
                                                                                                                                                            )
Pelletdensity(% theoretical), min                                            93                                                              )
3 Theoretical density, A10        23, g/cm                                    3.94 Theoretical density, B 4C, g/cm3                                            2.52 Clad material                                                                Zircaloy-4 Clad lu, inches                                                              0.332                                                            '
Clad 00, inches                                                              0.382 Cladthickness,(nominal), inches                                              0.025 Diametral gap, (cold, nominal), inches                                      0.025                                      ,
Active length, inches                                                        136.0                                                            ,
Plenum length, inches                                                        11.090 l
 
l 1
CESSAR tiinnem.                                                                                                                        i i
TABLE 4.21(Cont'd) 1 (Sheet 6of6)
MECHANICAL DESIGN PARAMETERS i
turnable Poison Rod (Cont'd)
Gadolinia - Urania Poison Rod Desian Absorber material                                                                        UO -Gd 0 2                23                    ,
Pellet diameter, inches                                                                  0.325 Pellet length, inches, min                                                              0.390 Pellet density (% theoretical), min                                                      94.75 Theoretical density, U02, g/cm                                                          10.96 B
3 Theoretical density, Gd23                    0 , g/cm                                  7.41 Clad materisi                                                                            Zircaloy-4                                    ;
Clad ID, inches                                                                          0.332                                          ,
Clad OD, inches                                                                          0.382 Clad thickness, (nominal), inches                                                        0.025 Diametral gap, (cold, nominal), inches                                                  0.007                                          .
      - Active length, inches                                                                  135 Plenum length, inches                                                                    9.527 Amendment B March 31, 1988
 
CESSAR ELui..
jQ TABLE 4.2-2
    ?
TENSILE TEST RESULTS ON IRRADIATED SAXTON CORE 111 CLADDING (REFERENCE 61) 21 Fluence (>l NeV) 4.7 x 10      n/ cat (estimated)
Uniform Strain    Totcl Location                                    Ultimate        In 2-in. Strain from    Testing      0.2% Yield            Tensile        6 age  In 2-in.
Rod  Botton    Ton            Str                Stre            Length    Sage (in.)-  'f*Fg      (1b/in.gssx 103 ) -(1b/in.ggthx 103 )    (%)  Lenoth-IR_
      ,c 80-  11-17      650'          61.4              65.6            2.2      f.8 B0    26-32      650            58.1              68.9            2.4    11.3 RD      3-9      650            62.2              70.0            2.0      4.2 RD    12;18    650.          60.5              65.4-            1.7      5.8 MQ-    12-18    675            70.4              77.4            1.9      6.1    -
MQ    28-34    675            66.0                75.1            1.6      6.2
          .'    FS    28-34    675            57.2                71.4            3.9    12.9
              -GL    12    675-          60.5        -
71.5            2.4      9.3 i
 
CESSAR Blutricam t                                                                                      _
TABLE 4.2-3
($heet1of2)-
                                              $-E POOLSIDE FUEL INSPECTION PROGRAM
 
==SUMMARY==
(REFERENCE 70)
Shutdown Reactor              Date/ Cycle              Insnection Procram Scone (")
Palisades            1973/lA        VE, GS, CS Maine Yankee          1974/1          VE, S, SRE, CS 1975/1A        VE, S 1977/2          VE, SRE 1980/4        VE, S, SRE 1987/9        VE, UT,'SRE Ft. Calhoun            1975/1        VE-1975/2        VE, CS 1977/3        VE 1978/4        VE, OM on DOE Test Bendles 1980/5        VE, DM on DOE Test Bundles 1981/6        VE, DM and SRE on DOE Test Bundles 1982/7        VE, DM and SRE on DOE Test Bundles            F St. Lucie-1          -1976/1        VE,-SRE 1978/1A        VE 1985/6        VE, UT, SRE 1987/7        VE, UT, SRE Calvert Cliffs-1      1976/1          VE, SRE on C-E/EPRI Test Bundles 1978/2          VE, SRE on C-E/EPRI Test Bundles 1973/3        VE, DM SRE on C-E/EPRI Test Bundles 1980/4        VE, DM SRE on C-E/EPRI Test Bundles 1982/5        VE, SRE on C-E/EPRI and C-E/BG&E Test Bundles 1983/6        VE, DM 1985/7        VE, DM, SRE on C-E/BG&E Test Bundles 1986/8        VE, DM, UT, SRE on C-E/GB&E Test Bundles 1988/9        VE, UT, SRE Calvert Cliffs-2      1984/5        VE, DM, S, SRE 1987/7        VE, UT, SRE 1989/8        VE, UT, SRE Yankee Rowe            1987/18        VE, UT, SRE Amendment F December 15, 1989
 
CESSARilWicuitu TABLE 4.2-J (Cont'd)
(Sheet 2of2)
C-EPOOLSIDEFUELINSPECTIONPROGRAM
 
==SUMMARY==
(REFERENCE 70)
Shutdown Reactor            Date/ Cycle                      Insnection Proaram Scone (a)
M111 stone-2        1977/1                VE 1982/4                VE St. Lucie-2          1987/3                VE, UT 1989/4                VE, UT, SRE ANO 2                1981/1-                VE, DM, SRE on C-E/EPRI Test Bundles 1982/2                VE, DM 1983/3                VE, DM, SRE on C-E/EPRI Test Bundles 1985/4                VE, DM, SRE on C-E/EPRI Test Bundles 1986/5                VE, DM, UT 1989/
San Onofre-2        1984/1                VE, DM 1985/2                VE, DM 1987/3                VE, UT, GS, SRE 1989/4                VE, UT, SRE, DM F
                    -San Onofre-3        1985/1                VE, UT 1988/3                VE, UT, SRE Palo Verde-1        1987/1                VE, DM
            ,                              1989/2                VE, DM Palo Verde-2        1988/l                YE, DM Waterford-3          1988/1-2              VE, UT, SRE (a) VE    Visual Examination GS    Gamma-Scanning CS    Crud Sampling S    Sipping UT    Ultrasonic Testing SRE Disassembly and Single Rod Examinations DM    Dimensional Measurements Amendment F December 15, 1989
 
    ,,        m    .
1
:L                  GRIPPER COUPLING ---*                                                  ',                                                        J a            y                G HOLDDOWN                                                        _
SPRING :                                                      ]                      o                                ,
Nk            =      ,I
                                                                                            ,J g
,                            LOCKING                ,
                  -                    h  i h                  '
I j
i l
SERIAL No.
LOCATIONS h
                                                                                                                        -l      & PLANT ID No.s -
f                e SPIDER                  \              SYSSO j (A          p                                  .__      N l"                                                      END FITTING
                                        -L                    ._:    ,
PLUNGER' y w"
                                                                                            =
252.97
                                          =                          =                    l O
R NGTH      =                          =>                                /fpRI                                          '
CONTROL ROD-
                                      ^
                                                                                                    /                                                ,
ASSEMBLY      --                        -                      w                                                .
7              PLENUM                                  j h N STAINLESS s s            SPACER STEEL 244.63 j'
s 5      :  - SILVER. INDIUM CADMlUM s s            CYLINDRICAL SEGMENTS 5.5                                                        ,
2.2 s't s s v1
                                                              -          e.816 DI A.      s s s '
N s 148"                                                      "
ACTIVE LENGTH                                                      l) s.s N s N s N s N
N N            STAINLESS STEEL N              SPACER
[
1.5 y            o                                                                                                Amendment F December 15,1989            ,
                                              .                                                                                      Figure fd du                      FULL STRENGTH CONTROL ELEMENT ASSEMBLY (4 ELEMENT)                                      4.2 3 a <
 
                                                                                                                                              ' ' ' ' " ~ - ~ ~ ~
j ,
Q-iEi!
GRIPPER                                                  @
COUPLING --    Oa                                                        o                  o 1
l HOLDDOWN                  m    4 SPRING %-                    '
l._- ,l j'            Q        ;i #          D LOCKING-        ,
NUT %    -
I              '"                                      i              J SERIAL
                                                                                                                                            -LETTER SPIDER --      k'                                                          ,
LOCATION (s aW                                  F                          sysso
                                                                                                                                                  & PLANT ID No.s 252.97 U          ?
L_Uy                                          _
PLUNGER %                        e PART.                t 244.63-            =                      =        STRENGTH                          h CONTROL SSEMBLY,                          y HOLDDO M u
SPRING Y                      v INCONEL.
SLUGS                  o 149 INCONEL
[
g REF.
                                                                                --      e.816 DI A.                  34, 4
                                        ,      y    .,        y                      y                          V j            ,
I Amendment F December 15,1989 Figurs            .
jg                              PART STRENGTH CONTROL ELEMENT ASSEMBLY 4.2 5
 
(
1
  +                                ,                                                                                                            __
s <<
m r,r                    , <<              . ca          . ca                o es                    s    *'s            s b
q l
                            &(,                                                                                                                          %
6 $,.)  .\
                                                                  /-m.,                                                  . . -m,          -
i f
I                                        f                  I
                            $ r                                                                                                y                        i
: h.                      l            (    I                k.s , ../                                                \ . , , ..
                                                                                                                          .                            ~ %,  ,IJ          .
9 ip                                                                                                                \),        l
: e.                                        ,-                                                                                                          ,a l%(              1                                                                                                                I    ,l
                                                                                                      . m    '
I P                                                      i l
                                                                                                        ~
(l                                                        \.        /
w[/                                                                                                                                      l
                                                                                                      'v  ,
ob
                    ,      l
                                      .(
WELD LOCATIONS I
hg          ,
TOP & BOTTOM OF                                              i.
l                                                                                                                      l 7              _=
7 s                              SPACER GRID                ,                            )
1; _ \          P;\)_\f/*                                                                (                    ,
rc          ^
y    c;                    *                                                ~
                                                                                                                                              ,            e  4
                                                  =    -
f l ;        i f      $ -h                    l                                                                            \
                                                                                                                                                                >        l dE                    sj e%
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s o'                  4. 3'      NUCLEAR DESIGN 4.3.1          DESIGN BASES                                                                l 1
The bases for the nuclear . design of the fuel and reactivity                              I control systems are discussed in the following paragraphs.                                I 4.3.1.1          Excess Reactivity and Fuel Burnun The . excess- reactivity provided for each cycle is based on the
                  . depletion characteristics of the fuel and burnable poison and on
  ,'              the ' desired burnup for each cycle.            The desired burnup is based on; an economic analysis of the fuel cost and the projected operating  load cycle fo:: System 80.          The average burnup is chosen to ensure  that the peak burnup is within the limits discussed in This design basis, along with the design                F Paragraph  4.2.3.2.10.
basis -in  Paragraph 4. 3.1. 8, satisfies General Design Criterion
: 10. .                                                                                    ,
                  '4.3.1.2          core Desian Lifetime and Fuel Renlacement Procram The-core design lifetime and fuel replacement program presented are based on a refueling interval of approximately.18 months with                    p more than one-third of the fuel assemblies replaced at. each refueling- in later cycles.          The typical 18-month refueling
                  - program presented replaces 108 fuel assemblies in the second cycle and 96 fuel assemblies'in subsequent cycles.
4.3.1.3          Nocative Reactivity Feedback
                  .In the power operating range,            the not effect of the. prompt inherent nuclear feedback characteristics                  (fuel temperature coefficient, moderator temperature coefficient, and moderator pressure coefficient) tends to compensate for a rapid increase in reactivity.      The negative reactivity feedback provided by the design satisfies General Design Criterion 11.
4.3.1.4            Reactivity Coefficients The values of each coefficient of reactivity are consistent with the design basis for net reactivity feedback (Paragraph 4.3.1.3),
and analyses that predict acceptable consequences of postulated accidents and anticipated operational occurrences (Acos), where such analyses include the response of the reactor protective system (RPS).
Amendment F 4.3-1                          December 15, 1989
 
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    '4.3.1.5          Burnable Poison Recruirements The burnable poison reactivity worth provided in the design is          '
sufficient    to~ ensure that the moderator coefficients' of
    ' reactivity are consistent with the design bases in Paragraph 4.3.1.4.
4.3.1.6        stability criteria Tha reactor and the instrumentation and control systems are              ,
designed to detect and' suppress xenon-induced power distribution        '
oscillations that could, if not suppressed, result in conditions ethat exceed the specified acceptable fuel design limits (SAFDLs).
The design of the reactor and associated systems precludes the possibility. of power level oscillations. This basis satisfies General Design Criterion 12.
4.3.1.7        Maximum controlled Reactivity Insertiesn Rate The core, - control element assemblies (CEAs), reactor regulating            '
system, and boron charging portion of the chemical and volume
      - control system (CVCS) are designed so that the potential amount and rate of reactivity insertion due to normal operation and                !
postulated reactivity accidents do not result in:
A. Violation of the specified acceptable fuel design limits                i (SAFDLs)
B. Damage to the reactor coolant pressure boundary (PCPB)
C. Disruption of the core or other reactor internals sufficient to impair the effectiveness of safety injection.
This design basis,      along with Paragraph    4.3.1.11,  satisfies General Design Criteria 25 and 28.
4.3.1.8        Power Distribution Control The  core power distribution is controlled such that,              in conjunction with other core operating parameters, the power distribution does not result in violation of the limiting conditions    for  operation  (LCOs). Limiting conditions    for operation and limiting safety system settings (LSSSs) are based on the accident analyses described in Chapters 6 and 15 such that specified acceptable fuel design limits and other criteria are not exceeded for accidents.        This basis, along with Paragraph 4.3.1.2, satisfies General Design Criterion 10.
4.3-2 l
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0 The COLSS, described.in Section 7.7 and Reference 1,      continually generates an assessment of the margin to linear heat rate and DNBR operating limits. The data required for these assessments include measured in-core neutron flux data, CEA positions, and coolant inlet temperature, pressure, and flow.        In the event of-an alarm indicating that an operating limit has been exceeded, power must be reduced unless the alarm can be cleared by improving either the power distribution or another process parameter.
periodically.The accuracy of the COLss calculations are verified          lF In addition to- the monitoring performed by COLSS, the RPS Core        B Protection Calculators (CPC, see Section 7.2) continually infer the core power distribution and DNBR by processing reactor coolant data, signals from ex-core neutron flux detectors, each containing three axially stacked elements,          and input from redundant reed switch assemblies to indicate CEA position. In the event the power. distributions or other parameters are perturbed as the result of an anticipated operational occurrence            -
that weuld violate fuel design limits, the high local power
* density or low DNBR-trips in the RPS will initiate a reactor trip.
4.3.2.2.2        Nuclear Design Limits on the Power Distribution The design limits on the power distribution stated here were employed during the design process both as design input and as initial conditions for accident analyses described in Chapters 6 and 15. However, for the monitoring system, it is the final O
operating limit determination that is used to assure that the consequences of an anticipated operational occurrence or postulated accident will. not be any more severe than the consequences shown in Chapters 6 and 15. The initial conditions used in this operating limit determination are actually stated in terms of peak linear heat generation rate and required power margin for minimum DNBR.
The design limits on power distribution are as follows:
A. Tge limiting three-dimensional heat flux peaking                  p F q, was established for full power conditions at 2.28.factog[q F
is defined in Section 4.4.2.2.2.1.C and is termed the nuclear power factor or the total nuclear peaking factor.
n  of 2.28 in combination with uncertainties and An  Fq allowances on heat flux which give the initial peak linear heat rate assumed in the safety analyses constitute one limiting combination of parameters for full power operation in the first cycle. Other combinations of parameters which will result in acceptable consequgnces for the safety analysis do exist, e.g., a higher F q is acceptable at a Amendment F 4.3-5                  December 15, 1989
 
CESSAR tilLm.
l L                  reduced.. power            level. Implementation      in  the  technical
                  . specification is via an operating limit on the monitored peak linear heat generation rate.
The thermal margin to a minimum DNBR of 1.24                                  F B.                                                                    (using the 1
C-E-1 CHF correlation as discussed in Sections                4.4.2.2. and 4 . 4 . 4 .1) ',  which is available to accommodate anticipated operational            . occurrences,      is  a- function    of    several parameters,. including.
: 1)          the coolant conditions L                    ii)            the axial power distribution lii)            the axiglly integrated radial peaking factor, F"r;
  ,                                  where F r is the rod radial nuclear factor or the rod radial peaking factor and is defined                  in
        ^
Section 4.4.2.2.2.1.A (referred to as F# in that                '
section).
The coolant conditions assumed in the safety analyses, an F"r of 1.55,    and the set of. axial shapes displayed in Figure                    4.4-3
            -constitute a set of limiting combinations of parameters 'r full power operation.            Other combinations giving acceptable          :cident
: 1.            analysis consequences are equally appropriate. Implementuuon of these limits in the technical specifications is via a power                            ,
operating limit based on DNBR which maintains an appropriate                        B '
amount of thermal margin to the DNBR limit.                  It will be shown in the following paragraphs that operation within these                        design limits is achievable.
4.3.2.2.3              Expected Power Distributions Figures 4.3-3 through 4.3-17 and 4.3-18 through 4.3-23 show typical first cycle planar radial and unrodded core average axial power distributions, respectively.                  They illustrate conditions expected at full power for various times in the fuel cycle as specified on the figures.              It is expected that for normal, base load operation of the plant, the operation of the reactor will be with    limited        CEA      insertion    so that the unrodded power distributions in Figures 4.3-3 through 4.3-23 represent the                          B expected power distribution during most of the cycle.                      If the plant is required to perform load follow operations, such as planned power level changes, the normal operation of the reactor may include full insertion of the lead part-strength CEA group.
Therefore,        Figures 4.3-3 through 4.3-17 show radial power distributions for both unrodded operation and for steady-state operation with the lead part-strength CEA group fully inserted.
It can be seen by these figures that the part-strength CEA group insertion has              only    a  small      effect  on tge radial power distributions and the radial peaking factors (F r) for different Amendment F 4.3-6                  Dece.aber 15, 1989
 
                                                                                                                                                                )
L    CESSAR tiiMem l                                                                                                                                                              .
1
!                                                                                                                                                              )
L    times in cycle.                Also,              the effect of full insertion of the                                                                    I l    part-strength CEA group on the axial peaking factor is negligible                                                                      g for steagy state operation.                                      The three-dimensional                                    peaking factor, F q, expected during steady-state operation g then just the  product of the planar radial peaking factor (F ry and the axial peaking factor. The maximum expected value of F q is 1.88                                                                                            ,
during the first cycle and,                                  as can be seen from the above                                                                l figures,            occurs near the beginning-of-cycle for- steady-state, base-loadedoperationwithnofull-strengthorpart-strengthCEAlB insertion.
Figures 4.3-24 through . 4. 3-35 show typical fuel cycle loading patterns, initial burnup distributions, and planar radial power distributions for the second .and third cycles based on a refueling interval of approxinately 18-months.                                                                      The expected power distributions for these cycles are similar to those of the                                                                      F first cycle except. for reduced power in fuel assemblies located on the periphery of the core and consequently higher radial peaking factors in the interior region of the core.                                                                  The expected power distributions are well within the nuclear design limits                                                                                          '
described in Section 4.3.2.2.2.                                  The uncertair.ty associated with these calculated power distributions is discussed in Section
                                                                                                                                                              ~
4.3.3.1.2.2.6.
The capability of the core to follow load transients without exceeding power distribution limitations depends                                                      on the margin to                                  1 operating limits compared to the margin required                                                    for base loaded,                                    t unrodded operation.                        In order to illustrate the                                core maneuvering                F                !
capability,            the        results                of          calculations                      of            the      power distributions and power peaking factors during load following transients - are discussed below.                                      The axial power distributions are calculated by VISIONS (Reference 2), a three-dimensional neutron diffusion code that considers the effects of the temporal and spatial variations of xenon and iodine concentration, CEA positions,              fuel            temperature                      and      moderator                          temperature distributions, soluble bgron congentration, and burnup.                                                                          The nuclear peaking factors F q and F r are synthesized in VISIONS using      the        calculated                    three-dimensional                  coarse-mesh                          power  8 distribution and input pin-to-box factors from MC (see Section 4.3.3.1.1.3).            Figures 4.3-36 through 4.3-39 show the calculated axial power distributions and associated nuclear peaking factors during a typical day of a maneuvering transient to 50 percent of the full power conditions.                                Figures 4.3-36 and 4.3-37, which represent maneuvering transients near beginning-of-cycle and end-of-cycle,              respectively,                    also              show  the                        locations      of full-strength and part-strength CEA groups during the transients.
The transients begin with the lead part-strength CEA group fully            inserted.                  Throughout                the        calculation                            of    the Amendment F 4.3-7                                            December 15, 1989
  -,    ,    ~ _ _ - _ _ _        _ _ _ . _ . , _                . . _ _ _ _ _ _ _            _ _ _ _ _ _ _ . _ _ _ _ _                    _- _ _ . .
 
CESSARanh o                                                                                        !
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power distribution during these transients it is assumed that the                                  I part-strength CEA groups are available for control of the axial                                    ;
power distribution. The two part-strength CEA groups are moved                                  )
to positions that minimize the difference between the current                                      '
B            '
shape index and the reference value of shape index that existed prior to the initiation of the maneuver.                In addition, the                            l positions of the part-strength CEA . groups supplement reactivity                                    l control provided by full-strength regulating rods, so that the calculated maneuvering transients can be accomplished without changing soluble boron concentration to compensate for reactivity changes due to power level and xenon.
The detailed radial' power distribution within any assembly is a                                    ;
function of the location of that assembly within the core as well as the time . in life, CEA insertion, and other considerations.
The. normalized assembly power distribution used for the sample DNB calculation discussed in Section 4.4.2.2. is shown on Figure 4.3-40. In Section 4.3.3.1. 2, the accuracy of calculations of the power distribution within a fuel assembly is discussed.
l    4.3.2.2.4        Allowances and Uncertainties on Power Distributions In cothparing the expected power distributions and implied peak linear heat generation rate (PLHGR) produced by analysis with the design limits stated in Section 4.3.2.2.2, consideration must be given to the uncertainty and allowances associated with on-line monitoring by COLSS.
The COLSS uncertainty analysis, as applied to System 80, is described in Section 7.7 and in Reference 1.                For monitoring linear heat rate, COLSS applies an overall uncertainty factor for 0
linear heat rate measurement, in addition to a power level uncertainty factor og 1.02. These factors are appyed to the COLSS monitoring of F q, such that a COLSS-measured F q of 2.28, the chosen design limit for F nq given in Section 4.3.2.2.2, will.
not result in exceeding the design limit for PLHGR.                                The allowances and uncertainties applied for the COLSS monitoring of thermal margin to the DNBR limit are also described in Section 7.7 and Reference 1.
4.3.2.2.5        Comparisons Between Limiting and Expected Power Distributions Ag discussed in Section 4.3.2.2.3, the maximum expected unrodded F q that occurs during the first cycle at full power is 1.88.
Augmenting this value by the required calcufational uncertainty                        B (Reference 3) provides an upper limit on F q of 1.98 which is well below the design target of 2.28.                  Additionally,              the calculations described in Section 4.3.2.2.3 show that, with Amendment B 4.3-8                    March 31, 1988
                                              ~. ..- .---                      - - . - -      - - . -
 
z Oi
:CESSARHEnce safety analysis is conservative since the lead regulating bank is not expected to be fully inserted at full power.          Similar CEA B
ejection event analyses are performed for zero power and several intermediate - powers. -The initial rod configuration assumed is the maximum transient insertion limit allowed by the Power Dependent Insertion Limit of the CEA groups at that power.              lF The CEA withdrawal incident from low power is analyzed with the maximum    calculated  differential    reactivity    insertion  rate resulting from a sequential CEA bank withdrawal with 40% overlap.
The CEA withdrawal incident from full power is analyzed from the insertion- of the lead bank which maximizes            the reactivity insertion and the power shape change during the CEA withdrawal.
Reactivity insertion rates are calculated by a static axial model of the System 80 core. The calculated reactivity insertion rate resulting from a sequential withdrawal of full-strength and part-strength regulating CEA groups is shown Figure 4.3-49. The          B calculated reactivity insertion rate for withdrawal of the lead full-strength regulating group is shown in Figure 4.3-50.                  =
The full-strength CEA drop incident is analyzed by selecting the              -
dropped CEA that maximizes the increase in the radial peaking factor. The radial peaking factors include an allowance for 15 minutes of xenon redistribution. A conservatively small negative reactivity insertion is used in the accident analysis.
The typical reactivity insertion during a          reactor scram    is presented in Section 15.1. This reactivity insertion is computed using axial models at various scram configurations, and it is                  '
used for all. accidents which are terminated by a scram, unless otherwise indicated.      The reactivity insertion is conservative since only the minimum shutdown worth of 10.0%ap is assumed to be available at hot full power. The scram reactivity insertion for the loss of flow event is implicit in the kinetic axial analysis.
4.3.2.6        criticality of Reactor Durina Refuelina The soluble boron concentrations during refueling are shown in Table 4.3-1. These concentrations ensure core during refueling does not exceed 0.95.
that the k"If of the 4.3.2.7        Stability 4.3.2.7.1        General Pressurized water reactors (PWRs) with negative overall power coefficients are      inherently  stable with respect to power oscillations. Therefore, this discussion will be limited to Amendment F 4.3-19                December 15, 1989
 
CESSAR Riinnem.
I xenon induced power distribution oscillations.                    Xenon induced'            ;
oscillations occur- as . a' result - of rapid perturbations to the power distribution which cause the xenon and iodine. distributions
      .to be out of phase with the : perturbed power distribution.                      This results ~ in a shift in the j.odine and xenon distribution that causes the power-distribution to change in an opposite direction from the initial. perturbation and thus an oscillatory condition is  established. The magnitude of. the power distribution oscillation can either increase or decrease with time. Thus, the core can be considered to be either unstable or stable with respect to these = oscillations.                Discussed below are the methods of: analyzing - the stability of the core with respect to xenon-oscillations. The tendency of certain types of oscillations to increase or to decrease is calculated, and the method of controlling unstable oscillations is presented.
4.3.2.7.2        Methods'of Analysis Xenon oscillations may be analyzed by two methods. The first method consists of an explicit analysis of -the spatial flux-distribution accounting for the space-time solution of the xenon concentration. Such a method is useful for testing various control strategies and evaluating transitional effects (such as .
power maneuvers).          The            second method . consists      of  modal perturbation theory. analysis, which is useful for the evaluation of the sensitivity of the stability to changes in the reactor
;        design characteristics, and for the determination of.the degree of stability for a particular oscillatory mode.
The stability of a reactor can be characterized by a stability index or a damping          factor which is defined as the natural exponent which describes the growing or decaying amplitude of the oscillation. A xenon' oscillation may                be  described    by      the following equation:
4 (E,t) = dg(E) + ad o(E)e bt sin (wt + 6) where 4(E,t)          is the space-time solution of the neutron flux, is the initial fundamental flux, do (E) 64 g (E)        is the perturbed flux mode, b                is the stability index, w                is the frequency of the oscillation, and 6                is a phase shift.
4.3-20
 
CESSAR WGem..                                                                      l i
l Modal analysis consists of an explicit solution of the stability                  '
index b using known fundamental and perturbed flux distributions.
L      A positive stability index b indicates an unstable core, and a negative value indicates stability for the oscillatory mode being investigated.      The stability index is generally expressed in units of inverse hours, so that a value of -0.01/h would mean
      .that the amplitude of each subsequent oscillation cycle decreases by about 25%'(for a period of.about 30 hours for each cycle).
Xenon. oscillation modes in PWRs can be classified into three general types: radial; azimuthal; and, axial.            To analyze the          i stability for each oscillation mode, only. the first overtone needs to be considered since higher harmonic modes decay more
  . rapidly than the first overtone.
4.3.2.7.3          Expected Stability Indices 4.3.2.7.3.1          Radial stability
                                                                                    +
A, radial xenon oscillation consists of a power shift inward and            V outward from the center of the core to the periphery. This                  e oscillatory mode is generally more stable than an azimuthal mode.
This effect is illustrated in Figure 4.3-51, which shows that for a bare cylinder the radial mode is more stable than the azimuthal mode. Discussion of the stability for radial oscillatory modes is therefore deferred to that for-the azimuthal mode.
4.3.2.7.3.2          Azimuthal Stability                                    i
                                                                                    ^
An azimuthal oscillation consists of an X-Y power shift from one side of the reactor to the other. Modal analysis for this type of oscillation is performed for a range of expected reactor operating conditions.
The expected variation of the stability index during the first cycle is shown in Figure 4.3-52.        These results are obtained from analyses which consider the spatial flux shape changes during the cycle,  the changes in the moderator and Doppler coefficient during the cycle, and the change in xenon and iodine fission yield due to plutonium buildup during the cycle.          As is shown on the figure, the expected stability index is no greater than
      -0.04/h at any time during the cycle for the expected mode of reactor operation.      Comparison of predicted stability index with those actually measured on operating cores,            as discussed in Section 4.3.3.2.3, provide a high confidence level in the prediction of azimuthal stability. Measurements of xenon spatial stability in large cores have been made          (Reference 6) which provide confidence in the methods that are used to predict the azimuthal stability of this core.
4.3-21
 
CESSAR nahm 4.3.2.7.3.3        Axial Stability An axial xenon oscillation consists of a power shift toward the top and bottom of the reactor core. This type of oscillation may be unstable during the first cycle.                                              Table 4.3-9 shows the calculated variation of the axial stability index during the first cycle.      It is anticipated that control                                            action with part-strength rods and/or full-strength rods may be required to lB limit the magnitude of the oscillation.                                          As discussed in Section 4.3.2.2, the axial power distribution is monitored by COLSS and the RPS. Based on the COLSS measurement of the axial power distribution, the operator may move either the full-strength or the part-strength CEAs so as to control any axial oscillations.
lB 4.3.2.7.4        Control of Axial Instabilities The control of axial oscillations during a power maneuver is illustrated in Figures 4.3-36 through 4.3-39. Part-strength CEAs                                            B (PSCEAs) are used throughout these maneuvers to limit the change in  the  power    distribution.                                        The  difference    between  an uncontrolled and a controlled xenon oscillation is illustrated in Figure 4.3-53.      It was assumed in the calculation of the controlled oscillation that the PSCEAs were moved in such a way l0 as to preserve the initial shape in the core prior to the initiating perturbation.                  The calculations are performed at the end of the first cycle, which corresponds to the expected least stable condition for axial xenon oscillations.
4.3.2.7.5        Summary of Special Features Required by Zenon Instability The RPS described in Section 7.2.2 is designed to prevent exceeding acceptable fuel design limits and to limit the consequences of postulated accidents.                                          In addition, a means is provided to assure that under all allowed operating modes, the state of the reactor is confined to conditions not more severe than the initial conditions assumed in the design and analysis of the protective system.
Since the reactor is predicted to be stable with respect to radial and azimuthal xenon oscillations, no special protective system features are needed to accommodate radial or azimuthal mode oscillations.      Nevertheless, a maximum quadrant tilt is prescribed along with prescribed operating restrictions in the                                              F event that the tilt is exceeded.                                            The azimuthal power tilt is determined by COLSS and included in the COLSS determination of core margin. The azimuthal power tilt limit is accounted for in the RPS.
Amendment F 4.3-22                    December 15, 1989
 
L  CESSARIHL m 4.3.3.3.4        Axial Xenon oscillations To check and confirm the predictions _of the linear modal analysis approach, numerical space-time calculations .were performed for both beginning and and-of-cycle.      The fuel and - poison burnup distributions- were obtained- by depletion with soluble boron control, so that the power distribution was strongly flattened.
Spatial Doppler feedback was included 'in- these _ calculations. In-Figure 4.3-62, the time variation of the power distribution along the core axis .is shewn' near end-of-cycle with reduced Doppler feedback.
The initial perturbation used to excite the oscillations was a 50% insertion into the top of the core of a'l.5% reactivity CEA bank fer-1 hour. The damping factor for this case was calculated to -be about 0.02 per hour; however, when corrected for finite-time step intervals by, the methods of Reference ' 23, the damping. f actor- is increased to . approximately +0. 04. When this    7 damping factor is plotted on Figure 4.3-63 at the appropriate          d eigenvalue _ separation for this- mode at and-of-cycle, it is          '~,
apparent -that- good agreement is obtained with the modified Randall-St. John distribution of the moderator coefficient about          '
the core midplane, and its consequent flux and adjoint weighted integrals of approximately zero.
Axial xenon' oscillation experiments performed at omaha at a core        ;
exposure of 7000 mwd /MTU and at Stade at beginning of cycle and at 12000 mwd /MTU (Reference 24) were analyzed with a space-time          5 one-dimensional axial model.      The results are given in Table
    .4.3-19 and show no systematic error between the experimental and analytical-results.
4.3.3.3        Reactor vessel Fluence calculation Model The method    for calculation of the maximum expected neutron fluence '(E>l MeV) to the reactor vessel over its design lifetime uses results obtained from two-dimensional transport theory          g calculations with the DOT code (Reference 16).        The DOT model uses the R-9 coordinate system to represent the geometry of the core, surrounding water, internals, and vessel.        The transport theory calculation of maximum local fluence to the vessel is based upon expected power history over the plant life.            The calculated vessel fluence includes an adjustment for observed differences between calculation end measurement based on analysis of surveillance capsule data for operating C-E plants (References 25-27), and additionally includes a +30% uncertainty factor.
Amendment B          =
4.3-41                March 31, 1988
 
    ,ii .
CESSARJinhn.
REFERENCE 8 FOR SECTION 4.3 C,
: 1.    " Overview                        Description .of            the        Core  Operating            Limit g Supervisory                              System    (COLSS),"        CEN-312,      Rev  01,1          C-E Proprietary Report, November 1986,
: 2. D. Bollacasa and' R. M. Versluis,- " VISIONS                                        -
Versatile, Interactive Simulator .of Nuclear Systems,"- Combustion                                                    E Engineering Paper. No. TIS-6964,                                          presented' at American Nuclear Society Winter Meeting, - November 29 - December. 3, 1981.
: 3.    "The-~ ROCS                      and DIT Computer Codes                    for Nuclear  Design,"                  ,
CENPD-266-P-A,-C-E Proprietary Topical Report, April 1983.
0
: 4. P. H. Gavin and P. C. Rohr, " Development and Verification of a Fuel Temperature Correlation for Power Feedback and Reactivity Coefficient Applicatien," Trans. Am. Nucl. Soc.,
30, 765 (1978).
,          5.    "C-E Fuel Evaluation Model Topical Report" CENPD-139, Rev.
01      (Non-Proprietary),                            CENPD-139          Supplement 1, Rev. 01              F (Non-Proprietary), July 1974,                                                                                      t
: 6. Krebs, W. D. and Brinkman, H., Proceedinas of Reaktortacunc 122ft, Dusseldorf, Germany, March 1976.
: 7. A. Jonsson,LJ. R. Rec, U. N. Singh, ." Verification of. a Fuel                                                    -l Assembly Spectrum Code Based on Integral Transport Theory"                                                            '
,                Trans. Am. Nucl. Soc. 28, 778 (1978)
: 8. Breen,        R.        J.,              et al.,  " HARMONY-System for Nuclear Reactor Depletion Computation," WAPD-TM-478, January 1965.
B
: 9. D. E. Kusner, et. al., "ETOG-1, A Fortran IV Program to Process Data from the ENDF/B. File to the MUFT, GAM and ANISN Formats,"-WCAP-3845-1, (ENDF-114), (1969).
: 10.  ~J. Adin, K. D. Lathrop. " Theory of Methods Used in the GGC-3
                                                              ~
Multigroup Cross Section Code." GA-715G, July 19, 1967.
l.
: 11. H. C.      Honeck, D.                        R. Finch, " FLANGE-II, A Code to Process Thermal          Neutron                      Data    from an          ENDF/B Tape,"    DP-1278, ENDF-152, (1971).
: 12. P. Kier, A. Robba, " RABBLE, A Program for Computation of Resonance                        Absorption          in    Multiregion        Reactor    Cells,"
ANO-7326, (1967).
I 1'
Amendment F 4.3-42                        December 15, 1989 l
4 o      ,
                      -n-..  - , -  . . , - - - , - .---.                        ,
 
                              .. .        .            -.        . . - . . . -    . . - - . ~ - - - .                - - - -        . . . _ - ~  . . . .      -
E il s : ! (-
    ,p-                                                                                                                                                              1 o
il se N                                                                  1                2      3                4 E                                                                                                    ~
BH          D0      St              D0 N = BOX No.                          5                6              7                8      9-              10                                1 f          F, = ASSEMBLY. TYPE gH                    00            CO          CO-      BH              D0 i
f
.I' (c                                            11-      12              13              14              15      16              17                                ,
BL    D0                    CO            D2          CO      D1              B H                          >
18        19        20              21            22              23
                                                                                                                                                                    ~
24              25                        2 11 Bt          D0      B                    U1            OL          D5      CL              D1 H
;;                                                                                                                                                            i.
26      27-        28              29            30              31      32              33 DO          CO      D1                    CO      -
D3          CL      D5              B H                    ,
34          36        36        37              38            39              40      41              42 B
H          CO          D2      BL                    D3          CO            D3      CL              D5                    -
                                                                                                                                                                    ?
43            44        45        46              47            48              49      50              51
            +
DO            CO          CO      D5                    CL            D3          BL      D5              BH                              '
o 52            53        54        55              56            57                58      59              60 Bt            B            D1      CL                    D5            CL          D5      B                D4 H                                                                            H 61            62
}                                            63        64              85              66              67      68              69 D0            D0          B      D1                  B              D5            B        D4              B H                            H                          H                        H Amendment F December 15,1999
                                                  .                                                                                      Figure SECOND CYCLE FUEL LOADING PATTERN
 
l FUEL      No. OF FUEL                                No. OF SHIM ASSEMBLY                                        NUMBER OF TYPE                                        ASSEMBLIES ENRICHMENT      RODS PER                                        RODS /'                                    gm B10/IN.
W/T% U235      ASSEMBLY                                  ASSEMBLY DO                                                    32              4.02                184 3.57                                                                0                                        .
52 1
4.02                  168                                                                                                      'I D1                                                    20                                                                                  16                                      0.022 3.57                  52 o                                D2                                                      8              4.02                  168 16                                      0.020 l                                                                                                          3.57                  52 D3                                                    16 3.57                168                                                                                                      ^
16                                      0.022 3.09                    52 l
                                                                                                            .57                172 D4                                                      4                                                                                  12                                      0.008 l                                                                                                        3.09                    52-l 3.57                172 D5                                                    28                                                                                  12                                        0.020 3.09                    52 XM                                                        MM                              XX                                                        l            XX X                                                          X                              XM                  E                        E                        EX XX                              XX                                                XX                                      XX
                                                                                                                                                                                                  ]
                                            ~~                                                        "                              "~                                                                    ~-
M                  Mt                                                  t XM                              MM                                        E        XXE                              EXX                    E
                ,                                                                      XM                ,                                                        XX M r7N                                                                          Xr1X M LJM                                                                          XL JX XX                                                                            XX XM                                XM                                        E        XXE                              EXX                    E g rax xr,
                                                            ' JM                            M' J
__      Er'X XL;X Xr-33                    _      _
XL JX XM                                MX                                  -            XX                                    XX X                                                          X                            XE                    E                      E                        EX XM                                                      XM                            'XX              i                                                      XX
          ,                                                                          D0                                                                  D1,D2,D3 3:::
rr
                                          ,, mi                                                    m v XX                                XX
                                                                    ,wm                  mXr, JN                      XL J E                        xX          y                                            WATER HOLE Xr,x N
xt;x XX          E g              LOWER ENRICHED FUEL PIN x                    x                                  0            HIGHER ENRICHED FUEL PIN N<EN        ;xE                    mx<    3 --
:CID XX                                XX IEX E            SHIM PIN
:cx                                                        -<v Amendment F D4,D5 December 15,1989 Figwe
(                                                        SECOND CYCLE ASSEMBLY FUEL LOADINGS
                          """""/'""''""                                                                    WATERHOLE AND SHIM PLACEMENT                                                                                  4.3 25
 
                ..g                              e                                                -
((
XXX- ASSEMBLY AVERAGE BURNUP (MWD /T) 4 BH      DO      BL      00 16757        0  17431        0 I
h        D0      CO      CO      By      D0 187311      0-    9944'  9940  18889      0 D0                                                              l BL                CO      D2      CO      D1      BH 18220      0        13505    0      12796    0      18832 SH        D1              D5      CL Bt                      D1 17262        0    18377        O  13874      0-CO      D3      CL      D5      BH 13622    0      17094    0      18853            i
              .                                                                                          i CO      D3    CL      D5 11293      0  17087      0 t
BL      D5        BH 18772    0      18899              !
BH        D4 18901        0 BH 18850 Amendment F December 15,1989
                                  .                                                        Figure gggg                      ASSEMBLY AVERAGE BURNUP DISTRIBUTION mFais        ==
4.3-26 AT BEGINNING OF SECOND CYCLE
 
a X.XX :
BOX AVERAGE (BOX POWER / CORE AVERAGE POWER)
X.XX  MAX. PIN FACTOR (MAX, PIN POWER / CORE AVERAGE POWER)
BH      DO      BL        DO 0.37    0.69    0.51        0.78 BH        D0      CO      CO      sg -      DO 0.42      0.93    0.92    0.96-    0.83        1.24 BL        00        CO      D2      CO      D1          By 0.46      1.11_      1.08    1.26    1.11      1.25        0.97 1.47 BH        D1      BL      D5      CL        D1 0.98      1.27    1.00    1'.29    1.17      1.27 CO      D3      CL      D5          By 1.12    1.23    1.12      1.25_      0.98 I
CO      D3        CL        D5 1.14    1.16    1.05        1.18 BL        D5          BH O.91    1.12        0.90 BH          D4 0.88      1.15 BH
(    0.90 l
Amendment F December 15,1989
                            .                                                            Figure PLANAR AVERAGE POWER DISTRUBUTION
_ jjfj                                                                                      l AT BEGINNING OF SECOND CYCLE, UNRODDED              4.3 27
 
o!-                                                                                                                                          y
                                                                                                                                                    'I X.X X ^
BOX AVERAGE (BOX POWERICORE. AVERAGE POWER) -
X.XX MAX. PIN FACTOR (MAX. PIN POWER / CORE AVERAGE POWER)                                                      l 7-
    .                                                                                                                                              4 B
H      00        8L        DO 0.34    0.63      0.48      0.69                  .
BH                00                    CO      CO        By        DO                      ,
0.38            0.81                  0.83  -0.87      0.78      1.08 8g      D0                CO                    D2'      CO        D1        BH 0.42      0.95            0.95                    1.24    1.05. 1.26      0.95 BH                                      E        D5        CL        D1
[31                      L O.91            1.29                    1.01    1.32    1.17    '1.36 1.47 CO                    D3      C.
t        D5        BH 1.12                  1.34'    1.15      1.33    -1.03 i
  ,                                                                                            CO      D3        CL        D5 l :-
1.19  1.32      1.12      1.30                  ,
Bt        D5        8H l'    .
1.01      1.26      0.99 BH        D4 0.98      1.28 BH 0.99 i
Amendment F December 15,1989 l;                                              m                                                                                  Figure I-i jfd                          hu                      PLANAR AVERAGE POWER DISTRIBUTION AT MIDDLE OF SECOND CYCLE, UNRODDED                              4.3-28
 
(
t
                    .X X BOX AVERAGE (BOX POWER / CORE AVERAGE POWER) .
                  - X.XX '
MAX. PIN FACTOR (MAX PIN POWER / CORE' AVERAGE POWER)
BH      D0                  D0-BL 0.41    0.88      0.53      0.73 4
BH        D0        CO      CO        BH        D0 0.42      0.64    0.88      0.90      0.81      1.09 BL        D0        CO        D2      CO        D1        BH 0.48      0.98      0.98      1.30    1.05      1.31      0.97 BH        D1        B        D5        CL        D1 L
0.94      1.35    1.02      1.29      1.13      1.38 1.48-CO        Q3      C.L        D5        BH 1.13    -1.34      1.10      1.27      0.99, CO      D3        CL        D5 1.14    1.30      1.05-      1.22 B          D5        By L
0.97      1.19      0.92 BH        D4 0.90      1.13 B
H 0.87 l
Amendment F December 15,1989
                                    .                                                              Figurs jfd M /
hu                DLANAR AVERAGE POWER DISTRIBUTION AT END OF SECOND CYCLE, UNRODDt!D                4.3 29 l
1 s-              .                          .- -
 
                                                                                                              ~    ' - ~ ~ - ~
yg                            ,
t;,      ,
                                                                                                                                  "'^; ;
r,                                                                                                                                      1 c ..
      'ffl
        .c l
l N:                                                    1          2        3          4                          l F                                                                                                            l CO        to      D0          D1                      ;
o.
.+,
N = BOX No.                      5            6      7        8        9          to
'~  -
                    . F = ASSEMBLY TYPE                      CO          E0      00 -      00      D3          E4 11      12            13    14        15      16        17 D2      E0          D1      E1      -D5      E3          DS 18          19      20            21    22        23      24        25 CO          EO      -CO          E1      CO        E2      D5          E6 26          27      26            29    30        31      32        33-EO-        D1'      E1          D1      E5-      D3      E2          D5 34        35-          36      37            38    39        40      41        42-
                          'E0        .D0          E1      CO          E5      D4        E5      CO          DO 43          44          45      46            47    48        49      50        51
['          .-    ,
E0          00          DS . E2          D3      E5        D5      E2          D2
!=
52          53          54-      55            56    57        58      59          60 D0        D3          E3.      D5          E2      CO        E2      D1          DO 61-        62          63      64            65    66 -      67      68          69                        -
l --
D1          E4          D5      E5          D5      00        D2      D0          BH p
1 Amendment F December 15,1989 m                                                              Figure gj                                          THIRD CYCLE FUEL LOADING PATTERN 4.3 30
= = _ _                                  -_                      ._-_          .    --        .      .  -                    .
 
ASSEMBLY                        NUMBER OF I
TYPE                    ASSE M S ENRLCHMENT            RODS PER          RODS /            gm B10llN.
W/T% U236            ASSEMBLY        ASSEMBLY E0                            24                4.03                144 0                  .
3.70                52 El                            14                4.03                168 16                  0.016 3.70                St Et                            24                3.70                168            16                  0.020 3.40                52 E3                              8              3.70                168 16                0.014 3.40                $2 E4                              4 4.03                17C' 8                0.012
* 70
                                                                          .                  St ES                            20                3.70                172              12                0.020 3.40                  52
                                                              ""                                                                        ~
EE xx an                                    XX E
                          ""
* d ~~
g    nm          E        E      EX      m-                            "Y XX              An                    XX              XX
                ~~
t            x                    ~~      ~
                                                                                                  ] ~~      ~~
                                                                                                                    . t            X
                                                                                                                                          ~~
                                                    "=
m    XXE            EXX    E                XXE        '
EXX
* X.X                                    aA gr,x                                      xr 3x                                gr ,g ELJ
                                            ==
K                                    XLJM                                  Kt JX XX                                    XX KR                  E A  nn          EXX    E                AAm          mam
                ~~                                          ~~      ~~                            ~~      ~~            ~
                                                                                                                                              ~
u                            x        X                        L              t MM                        nn                        na              XX                    XX              XX x                                                x  xE          E        E      ex      am                            EX ww                                    ww          _,
                                                                                                              --                          wv E0                                    Et,Et,E3                                    E4 00                                          EO KK                      MK
                --      lEj$ "                "
x CI      ""~
                            ,            ,,          m I          ur,x                                                  ] WATER HOLE x<;x B
1 XX          E                                        B LOWER ENRICHED FUEL PIN N
l <>_.
x          x          -~
D HIGHER ENRICHED FUEL PIN
                                                -at    J  N xx                        xX                                          E SHIM PIN an                                          XX E6                                                                            Amendment F December 15,1989 THIRD CYCLE ASSEMBLY FUEL LOADINGS 90MBesflee tsalNaamuse                                              WATERHOLE AND SHIM PLACEMENT 4.3 31 l
 
i l
l 1
l J
XXX                                    ASSEMBLY AVERAGE DURNUP (MWDIT)
                                                                                                                                                                                                                                                                                                                                    )
i CO                            E0                            DO                              D1 I
30847                              0                          10808                          21902                                        )
l CO                    E0      00                          00                            D3                            E4 1
29883                0      14084                        13589                          22227                                0 i
D1                          E0                    D1      E1                          DS                              E3                            05 21672                            0              22345        0                          22289                                0                      20846                                        6 l
l CO                    E1      CO                          Et                            DS                            ES                                                l 30549                  0    24069                              'O                        22271                                0                                      i t
01      E5                          D3                              Et                            DS                                                (
23199        0                        22782                                0                        21545 l-                                                                                                                                                                                      D4                          Es                            CO                              00 title                              0                      24795                          11877                                          ;
i                                                                                                                                                                                                                                                                                                                                    ,
06                            Et                              D2                                                !
20881                                  0                      21792 D1                              00 21916                            18144 B
H 19925 Arpendment F December 15,1989
                                                                                                                                      .                                                                                                                                                              Figure ASSEMBLY AVERAGE BURNUP DISTRIBUTION 1
MJ                                                                                                                        AT TsE .Eoinnino Or TwE TaiRD CYCtE                                                                                            4.3 32
 
X.XX DOX AVERAGE (DOX POWERICORE AVERAGE POWER)
X.XX MAX. PIN FACTOR (MAX. PIN POWER / CORE AVERAGE POWER)
CO      E0        D0        D1 0.30    0.70      0.83      0.58 CO        E0        00      00        D3        E4 0.35      0.85      0.88  0.97      0.90      1.15 D2      E0        D1        El      D5        E3        D5 0.52  1.04      1.03      1.20  0.99      1.17      1.03 CO        E1        CO      Et        D5        El 0.82      1.23      0.94    1.18      1.03'    1.24 i
;                                                        D1        El      D3        E2        DS 1.15      1.29    1.08      1.22      1.03 D4      El        CO        D0 1.15    1.32      1.01      1.29 D5        E2        D2 1.16      1.54      1.28 1.83 D1        D0 1.27      1.25 8H 1.07 Amendment F December 15,1989
                                          .                                                        Il Figure
                                      ""      PLANAR AVERAGE POWER DISTRIBUTION AT THE BEGINNING OF THE THIRD CYCLE, UNRODDED                    4333
 
      .            .      ._. _ _              ._ _ . _                    . ~.. _ _        _ .._. _ _ _ _.                              _ _ . __... _. . _. .. _                              ._. _        . _ _ _ _ . _ . _ _ . . _ _
                                                                                                                                                                                                                                              )
1 l
1 X.XX BOX AVERAGE (BOX POWERICORE AVERAGE POWER)
X.X X MAX. PIN FACTOR (MAX. PIN POWER / CORE AVERAGE POWER)
                                                                                                                                                                                                                                              ]
                                                                                                                                                                                                                                            -l
                                                                                                                                                                                                                                              )
l CO                          E0                            D0          D1                                      >
0.30-                        0.67                        0.43          0.59                                i CO                    E0            00                        00                              D3          E4
                                                                                                                                                                                                                                              )
0.34                      0.80      0.84                        0.94                        0.91          1.14 f
D2              E0                      01          El                          D5                            E3          05                                      !
0.49            0.95                      1.00        1.27                      1.01                        1.28          1.07                                i
!                                                                                                                                                                                                                                              i CO                    El            CO                        Et                              D5          ES                                      '
O.81                      1.31      0.94                        1'.30                        1.08        1.32                                i i
D1            El                        D3                              Et          D5                                      !
1.16        1.34                        1.11                        1.31          1.08 1.49                                                                                                      I D4                        Es                              CO          D0 l
1.13                        1.34                        0.98          1.14                                !
l D5                              Et          D2 1.11                        1,32          1.14
                                                                                                                                                                                                                                              +
r D1          D0 1.12        1.05 i
I N
                                                                                            ,                                                                                                              0.89 l
l Amendment F l
December 15,1989
                                                                            .                                                                                                                                Figure PLANAR AVERAGE POWER DISTRIBUTION JFJ                                                                    AT THE MIDDLE OF THE THIRD CYCLE, UNRODDED                                                                          4,3 34
      . . . , , . , , - . ,    .,..,_c..    .,,--..,.m,...,..,,,.,.,n,,,.                          y,,.      ,,,,m.m.m.,mm.  ,,,..w.,,,.,                      ,-.__. _ _ _ _ _ ___ _
 
e                                                                                        ;--
t X.X X DOX AVERAGE (DOX POWERICORE AVERAGE POWER)
X.X X MAX. PIN FACTOR (MAX. PW POWER / CORE AVERAGE POWER)
CO      E0      00          D1 0.34    0.70    0.46      0.43 CO        E0      00      00      DS-        E4 0.37      0.41    0.84    0.95    0.92      1.16 D2      E0        01      E1        DE      E3        DS 0.52    0.95      1.01    1.30      1.03    1.29      1.04 CO        El      CO        Et      D5        E5 0.43      1.32    1.00      1.34    1.08      1.32 1.47 D1      E5      D3      E2        DS 1.14    1.32    1.09    1.32      1.05 04      E6      CO        00 1.04    1.30    0.95        1.10 05      E2        D2 1.06    1.28      1.07 D1        D0 1.04      0.97            l OH 0.83 Amendment F December 15,1989 m                                                            Figure
      $fd        hw PLANAR AVERAGE POWER DISTRIBUTION AT THE END OF THE THIRD CYCLE, UNRODDED            4.3 35
 
l C E S S A R tl W icar..                                                                                                                                    4 4.4                              TEERMAL AND EYDRAULIC DESIGN This section presents the steady-state thermal and hydraulic                                                                                                !
analysis of the reactor core, the analytical methods, and the experimental work done to support the analytical techniques.
Discussions                          of  the  analyses              of                anticipated            operational                                i occurrences and accidents are presented in Chapter 15. The prime                                                                                            I objective of the thermal and hydraulic design of the reactor is to ensure that the core can meet steady-state and transient performance requirements without violating the design bases.
4.4.1                              DESIGN BASES Avoidance of thermally or hydraulically induced fuel damage during normal steady-state operation and during anticipated                                                                                                  ,
operational occurrences is the principal thermal hydraulic design                                                                                          j basis.                          The design bases for accidents are specified in Chapter                                                                      !
: 15.                In order to - satisfy the design basis for steady-state operation and anticipated operational occurrences, the following                                                                                            i design limits are established, but violation of these will not                                                                                              >
necessarily result in fuel damage. The reactor protective system                                                                                            ,
(RPS) provides for automatic reactor trip or other corrective action before these design limits are violated.
4.4.1.1                              Minimum Denarture from wuelente soiline Ratio (DNBR) l The minimum DNBR shall be such as to provide at least a 95%
probability with 95% confidence that departure from nucleate boiling (DNB) does not occur on a fuel rod having that minimum DNBR during steady-state operation and anticipated operational occurrences.                          A value of 1.24 using the CE-1 correlation coupled                                                  F with the CETOP code provides at least this probability and confidence.
4.4.1.2                              Evdraulie stability Operating conditions shall not lead to flow instability during steady-state operation or anticipated operational occurrences.
4.4.1.3                              Fuel Desian Bases i    A.            The peak temperature of the fuel shall be less than the melting point (refer to Sections 4. 2.1. 2. 4. 4 and 4. 2.1. 3. 4 )
during steady-state operation and anticipated operational                                                                  g occurrences.
Amendment F 4.4-1                                  December 15, 1989
 
CESSARtume.                        .
I l
B. The fuel design bases for fuel clad integrity and fuel assembly integrity are given in section 4'.2.1. Thermal and                                            I hydraulic parameters that influence the fuel integrity                                                '
include maximum linear heat rate,                          core coolant velocity,                      l coolant temperature, clad temperature,                              fuel-to-clad gap                )
conductance, fuel burnup, and UO temperature. Other than                                            '
the design limits already specin;ed,                            no limits need to be applied to these parameters directly.                            Conformance with the              :
I design limits specified here,                      and conformance with            the design bases specified in section 4.2.1., are sufficient to ensure fuel clad integrity, fuel assembly integrity, and the                                        '
avoidance of thermally or hydraulically induced fuel damage for ' steady-state operation and anticipated occurrences of                                            1 moderate frequency.                                                                                  ! l 4.4.1.4              Coolant Flow, Velocity, and Void Fraction
                                                                                                                  )
  .. The - primary coolant flow with all four pumps in operation shall                                            ;
be neither less than the design minimum nor greater than the                                              :
oesign maximum. A percentage of the flow entering the reactor                                              i vessel is not effective for cooling the core. This percentage is called the core bypass flow.                  The design minimum value for the calculated core flow is obtained by subtracting the design                                                !
maximum value for the calculated core bypass flow from the design                                            l minimum primary coolant flow.                  For thermal margin analyses, the                              i design minimum value ,for the calculated core flow is used. The                                            i design minimum primary coolant flow is listed in Table 4.4-1.
The design pre-core and post-core maximum primary coolant flows                                            ;
are equal to 1.22 and 1.16 times the design minimum. The design maximum primary coolant flow is used in the determination of design hydraulic loads in the manner described in Section 4.4.2.6.3.
Design of the reactor internals provides that the coolant flow is distributed to the core such that the core is adequately cooled during          steady-state          operation    and    anticipated          operational occurrences. Therefore, no specific orificing configuration is used.
Although the coolant velocity, its distribution, and the coolant voids affect the thermal margin, design limits need not be applied to these parameters because they are not in themselves limiting.          These parameters are included in the thermal margin analyses and thus affect the thermal margin to the design limits.                                          j i
                                                                                                                    )
1 4.4-2
 
CESSARtuh m.
                                                                                                                                      )
4.
 
==4.2      DESCRIPTION==
OF TIERMAL AND EYDRAULIC DESIGN OF THE                                                                  l REACTOR CORE 4.4.2.1        summary Comparison                                                                                                I The thermal and hydraulic parameters for the reactor are listed in Table 4.4-1. A comparison of these parameters with those for                                                                1 System 80 (Docket No. STN-50-470F) and the Waterford steam                                                    8 Electric Station Unit 3 (Docket No. 50-382) are included in this                                                                j table.                                                                                                                          ;
1 The only significant difference between System 80+ and System 80 thermal and    hydraulic design is the reactor inlet coolant temperature. The principal differences between the System 80 and Waterford designs are the total core heat output and reactor coolant inlet temperature.
4.4.2.2        Critical Meat Flux Ratios                                                                                  ~
4.4.2.2.1        Departure from Nucleate Boiling Ratio                                                                    (
The margin to DNB in the core is expressed in terms of the departure from nucleate boiling ratio (DNBR).                                              The DNBR is defined as the ratio of the heat flux required to produce departure from nucleate boiling at the calculated local coolant                                                                .
conditions to the actual local heat flux.
    .The DNB correlation used for design of the core is the CE-1                                                                  l correlation (1) (2) . Based on                      statisticalisevaluation                of the CE-1 concluded -that the correlation  and      relevant          data,                      it appropriate minimum DNBR is 1.20 (3).                                  The design minimum DNBR has increased with the application of Statistical Combination of                                              B Uncertainties (SCU) methods.                              Engineering factors, rod pitch, bowing and clad diameter factors will be combined with other uncertainty factors at the 95/95 confidence / probability level and                                          F it is expected to yield a higher design limit of 1.24 on CE-1 minimum DNBR. This limit is then used in conjunction with a                                                                ,
!    CETOP model based on nominal dimensions (See Section 4-.4.2.9.5).
Table 4.4-1 gives the value of minimum DNBR for the coolant conditions and engineering factors in the table, for the radial power distributions in Figures 4.4-1 and 4.4-2, and for the 1.26 peaked axial power distribution in Figure                                      4.4-3.        Values of l    minimum DNBR or maximum fuel temperature at the design overpower cannot be provided with any meaning.                                    The concept of a overpower is not applicable for the System 80+ cores since                                          design the lf reactor protective system prevents the design basis limits from being exceeded.
l l                                                                                        Amendment F 4.4-3                              tecember 15, 1989
 
CESSAR ttWicamn A comparison of the minimum DNBRs                                                      computed using different correlations for the same power, flow, coolant temperature and pressure, and power distribution is presented in Table 4.4-2.
The minimum DNBR values in both the limiting matrix subchannel and the limiting subchannel next to the guide tube are presented.
The. correlations compared are the CE-1 correlation, the original W-3 correlation (4), and the revised W-3 correlation (5).                                                        The differences between the original and revised W-3 correlations as used here are in the C-f actor . and the cold wall correction factor.
Additional comparisons are contained in CENPD-162-A (1).                                                          In general, the CE-1 correlation tends to predict lower values of CHF with high inlet subcooling and higher values of CHF with low inlet subcooling,                                                                                                      g The TORC and the CETOP computer codes (1) (2) are used to compute the local coolant conditions in the core and thereby the minimum DNBR.                  A discussion  of the CE-1 DNB correlation and the analytical                methods is presented in Sections 4.4.4.1 and 4.4.4.5.2, respectively.
4.4.2.2.2                    Application of Power Distribution and Engineering Factors Distribution of power in the core is expressed in terms of                                                                      !
factors that define the local power per unit length produced by the fuel relative to the core average power per unit length produced by the fuel.                The method used to compute these factors, which describe the core power distribution, is discussed in Section 4.3. The energy produced in the fuel deposits in the j
fuel pellets, fuel cladding, and the moderator and results in the generation of heat in those places.                                                      The fraction of energy l
deposited in the fuel pellet and cladding is called the fuel rod energy deposition fraction.                                Accordingly, the core average heat flux from the fuel rods is determined by multiplying the core power by the average fuel rod energy deposition fraction and then dividing by the total heat transfer area. The energy deposition fractions used for DNB analyses for the average and the hot fuel rods are given in Table 4.4-1.
The effects on local heat flux and subchannel enthalpy rise of deviations from nominal dimensions and specifications within tolerance are included in thermal margin analyses by certain factors called engineering factors. These factors are applied to increase the local heat flux at the location of minimum DNBR and to increase the enthalpy rise in the sub-channel adjacent to the Amendment B 4.4-4                                March 31, 1968
 
j f
CESSAR timb .
i rod with the minimum DNBR.        Diversion crossflow and turbulent interchannel mixing are not input as factors on subchannel enthalpy rise but are explicitly treated in the TORC and CETOP            ,
B analytical models.
Uncertainties in the power distribution factors are discussed in Section 4.4.2.9.4.                                                        ,
Statistical Combination of Uncertainties          (SCU)  methods,  as  -
described in Reference 8, were used to statistically combine the        B uncertainties of the thermal hydraulic code input parameters (system. parameters) . This SCU methodology with plant-specific data    is  statistically combined with CE-1 CHF correlation f
statistics at the 95/95 confidence / probability level to yield an increased DNBR limit. This limit is approximately 1.24 when the following uncertainties are combined:                                    ,
a)    uncertainty in the inlet flow distribution; b)    systematic variation on fuel rod pitch;                            ,
c)    systematic variation on fuel clad OD; d)    engineering enthalpy rise factor; e)    engineering heat flux factor;                                      ,
f)    penalty on DNBR (minimum) due to fuel rod bowing; and, g)    statistics  associated  with  the  NRC-approved  1.19  DNBR limit (2).                                                      8 Also included in the MDNBR limit is the penalty due to the CHF correlation uncertainty and a 0.01 penalty for the HID grids, as i  well as penalties imposed by NRC to account forCHFcorrelationlf
    " prediction uncertainty" and TORC code uncertainty.
DNBR limit is used in safety analysis, CPC trip setpoints and The 1.24 COLSS power operating limit calculations in conjunction with a          B CETOP model based on a nominal geometry.
l 4.4.2.2.2.1          Power Distribution Factors l
A. Rod Radia7. Power Factor                                          4 The rod radial power factor is the ratio of the average
* power per unit length produced by a particular fuel rod to the average power per unit length produced by the average powered fuel rod in the core. The maximum rod radial power factor is the ratio of the average power per unit length produced by the highest powered rod in the core to the average power per unit length produced by the average powered fuel rod in the core.      Radial power distributions are dependent upon a variety of    parameters (e.g., control rod insertion, power level, fuel    exposure). The core wide and hot assembly radial power      distributions used for a typical DNB analysis are shown in Figures 4.4-1 and 4.4-2.
Amendment F 4.4-5                December 15, 1989
 
l CESSAR tHanc.n..                                                                                          !
I 4
The maximum rod radial power factor for those figures is selected as 1.55 for better comparison with System 80 and the Waterford Station Unit 3.                        The actual maximum rod radial lB power factor in the core will normally be lowert but, it is not limited to a maximum value of 1.55. The only limits are those specified in Section 4.4.1. The protective system in                                            !
conjunction with the reactor operator utilizing the core                                              j operating limit supervisory system (COLSS) ensures that                                              <
those design limits are not violated.                                                                  l B. Axial Power Factor i
The axial power factor is the ratio of the local power per                                            :
unit length produced by a fuel rod to the average power per                                            l unit length produced by the same fuel rod.                                    The maximum            i axial power factor is the ratio of the maximum local power per unit length produced by a rod to the average power per                                            4 unit length produced by the same fuel rod.                              The axial power              l
.';-        distribution directly affects DNBR.                                                                    i 1
Typically, the farther the location of the peak heat flux is                                          ;
from the core inlet, the lower the value of the peak heat                                            )
flux needed to reach the DNBR limit.                                On the other hand,                '
fuel temperature is almost independent of the location of                                            l the peak heat flux and is principally dependent on the value                                          1 t        of the peak heat flux or linear heat rate. The axial power distribution and the maximum rod radial power factor are                                              ;
[        continuously determined and processed through the COLSS and
  -        the RPS such that the design basis limits are not exceeded.
Section 4.3 describes the power distribution and its
                                                                                                                  )
control.        Figure  4.4-3            shows              several        axial      power distributions used for this analysis.                              The minimum DNBR in                l Table 4.4-1 is determined using the 1.26 peaked axial power                                            l distribution, whereas the maximum heat fluxes are determined                                          !
using the 1,47 peaked axial power distribution.
C. Nuclear Power Factor The nuclear power factor is the ratio of the maximum local power per unit length produced in the core to the average power per unit length produced by the average powered fuel rod in the core.        It is conservatively calculated as the product of the maximum axial and radial power factors.                                      For better comparisons with System 80 and Waterford Station Unit                                      B 3, a value of 2.28 is selected for computing maximum heat fluxes.      The actual value of the nuclear power factor will normally be lower throughout the cycle;                                but, it is not limited to a maximum value of 2.28. The design limits are those specified in Section 4.4.1.                                The protective and supervisory systems assure that those design limits are not violated.
Amendment B 4.4-6                                March 31, 1988
 
CESSARBRMem 4.4.2.2.3        Fuel Densification Effect on DNBR The perturbation in local heat flux due to fuel densification is given in Table 4.4-1.      As shown in CENPD-207(1) (see Section 4.4.4.1), much larger local heat flux variations have no significant adverse effect on DNB.          Therefore, no specific allowance is made or required for the effect on DNBR of local heat flux variations due to fuel densification.
4.4.3.3        Linear meat eeneration Raig The core average and maximum fuel rod linear heat generation rates are given in Table 4.4-1. The maximum fuel rod linear heat generation rate is determined by multiplying the core average fuel rod linear heat generation rate by the product of the nuclear power factor, the engineering factor on linear heat rate, and the ratio of the hot to the average fuel rod energy deposition fractions. The effects of fuel densification are not included in the maximum fuel rod linear heat generation rate            .
presented in Table 4.4-1; although, to determine the maximum            :
local linear heat generation rate including the effect of gaps occurring between the fuel pellets, the augmentation factor is          ,
applied.
4.4.2.4      void Fraction Distribution The core average void fraction and the maximum void fraction are calculated using the Maurer method (10). 0    The void fractions discussed below are values for the reactor operating conditions and engineering factors given in Table 4.4-1, for the radial power distribution in Figure 4.4-1 and 4.4-2, and for the 1.26 peaked axial power distribution in Figure 4.4-3.          For these conditions, only subcooled boiling occurs in the core.
The core average void fraction is essentially zero.        The local B maximum void fraction is 0.4% and occurs at the exit of the subchannel adjacent to the rod with the minimum DNBR.            The average exit void fractions and qualities in different regions of the core are shown in Figure 4.4-4 for the core radial power distribution shown in Figure 4.4-1.      The axial distribution of void fraction and quality in the subchannel adjacent to the rod with the minimum DNBR is shown in Figure 4.4-5. The average void fraction in that subchannel is less than 0.1%.                      B 4.4.2.5        core coolant Flow Distribution The core inlet flow distribution is required as input to the TORC thermal margin code (refer to Section 4.4.4.5.2). The inlet flow distributien for 4-loop operation was determined from a System 80 Amendment B 4.4-9                March 31, 1988
 
I i
CESSARtmVicmo                                                                                      1 l
                                                                                                          )
reactor flow model test.            Descriptions of the model test and the resulting      core        inlet      flow        distribution    are    given      in        3 Section 4.4.4.2.1.                                                                                l 4.4.2.6          ggIt, Pressure Drons and Evdraulie Loads                                        ;
4.4.2.6.1          Reactor Vessel Flow Distribution                                              !
The design minimum coolant flow entering the four reactor vessel inlet nozzles is given in Table 4.4-1.                        The main coolant flow path in the reactor vessel is down the annulus between the reactor vessel and the core support barrel, through the flow                                      ;
skirt, up through the core support region and the reactor core, through the fuel alignment plate, and out through the two reactor l      vessel _ outlet nozzles. A portion of this flow leaves the main l      flow path as shown schematically in Figure 4.4-6.                        Part of the i
bypass flow is used to cool the reactor internals in the areas not in the main coolant flow path and to cool the CEAs.                              Table 4.4-3 lists the bypass flow paths and the percent of the total vessel flow that enters and leaves these paths.
The thermal margin calculations conservatively use the design maximum bypass flow of 3.0% of the total vessel flow as compared to the calculated bypass flow of 2.3% shown in Table 4.4-3.
4.4.2.6.2          Reactor Vessel and Core Pressure Drops                                        ,
The irrecoverable pressure losses from the inlet to the outlet nozzles are calculated using standard loss coefficient methods and information from System 80 flow model tests. These pressure losses have been verified by results from the final flow test on the complete System 80 reactor flow model, and are further                                  y confirmed by operational data from Palo Verde Unit 1.
Pressure losses at 100% power, the design minimug primary coolant flow, and an operating pressure of 2250 lb/in. , are listed in Table 4.4-4 together with the coolant temperature used to calculate each pressure loss.                    The calculated pressure losses include both geometric and Reynolds number dependent effects.
4.4.2.6.3          Bydraulie Loads on Internal Components The significant steady state hydraulic loads which act on the reactor internals during post-core steady state operation are listed in Table 4.4-5.                      These loads are determined from analytical methods and from results of reactor flow model and component test programs                    (refer to Sections 4.4.4.2.1 and 4.4.4.2.2, respectively).              The design hydraulic loads consist of steady state drag and impingement loads, and the fluctuating l      loads  induced by pump-induced pressure pulsations, vortex l
shedding, and turbulence.            ,
l l'                                                                          Amendment F i-                                                  4.4-10                  December 15, 1989 l
 
1 CESSAR timnem.=                                                                                          '
i l
                                                                                                            ~
coefficients,                    diversion      crossflow        resistance  and momentum parameters, turbulent interchange constants, and hot fuel rod energy deposition fraction.
: 2. The uncertainty in the analytical model to compute the                                        l actual distribution of flow and the local subchannel                                          i coolant conditions.
                                                                                                            ]
: 3. The uncertainty in the CE-1 correlation to predict DNB.
B. The following paragraphs discuss the above uncertainties and the allowances for them, if needed, in the thermal margin                                            ,
analysis of the cores i
: 1. Uncertainty in the input to the core analytical models
(
: a.                                        core geometry, as manifested by Uncertainty                    in manufacturing                    variations      within        tolerances,    is considered by the inclusion of engineering factors in                                  .a the DNBR analyses; see Section 4.4.2.2.2 for discussion                                  i,. ,
of the method used to compute conservative values.                                          ;
: b. Uncertainties on the power distribution factors are applied in the COLSS and RPS (see Section 7.7).
: c. The core inlet flow distribution is obtained from flow                                    '
l model        testing                  discussed      in    Section        4.4.4.2.          .
Uncertainties in the core                            flow    distribution are l ,
included in the design method for TORC analyses.
: d. Uncertainties                      in    the      core    inlet        temperature distribution and core exit pressure distribution are included in the design method for TORC analyses.                                            ;
: s. The Blasius single-phase friction factor equation for smooth rods is given and shown to be valid in Section                                        .
4.4.4.2.3.                    The spacer grid loss coefficient for the high impact grid is obtained from pressure drop data                                        .
discussed in Section 4.4.4.2.3.
: f. The value of minimum DNBR is relatively insensitive to crossflow resistance and momentum parameters (1).
: g. Section 4.4.4.1 describes the testing to determine the inverse Peclet number which is indicative of the turbulent flow interchange between subchannels.                                The inverse Peclet number is input to the TORC code and is used to determine the effect of turbulent interchange on the enthalpy rise in adjacent subchannels.                            From the testing, a value of 0.0035 is justified.
4.4-15
 
l
                    .CESSAR tlNncanow
: h. The same fuel rod energy deposition fraction is used for the hot rod as for the average rod. The hotter the                                                                                  i rod, the lower is the actual value of energy deposition                                                                                  I fraction with respect to that for the average rod.                                                                            A lower energy deposition fraction reduces the hot rod-                                                                                    ;
heat flux and thereby increases its DNBR. The use of the average rod energy deposition fraction for the hot rod is therefore conservative.                                      See Section 4.3 for a                                            1 discussion of the calculation of the energy deposition fractions.
l
: 2. Uncertainty in the analytical model:
L The ability of the TORC code to predict accurately I
subchannel local conditions in rod bundles is described in CENPD-161                (1).                        The ability of the code to predict accurately the core wide coolant conditions is described in CENPD-206                (L3 ) .                      However, an allowance for TORC code uncertainty is included in the Statistical Combination of                                                                              B      j Uncertainties analysis as discussed in Section 4.4.2.9.5.                                                                                    ;
3.,  Uncertainty in the DNB correlations t
The uncertainty in the DNB                                      correlation is determined by a                                                i statistical analysis of DNB                                    test data. A value of 1.20 has                                        B
* been shown to provide a 95%                                    probability with 95% confidence that DNB will not occur on                                      a fuel rod having that minimum                                                ,
DNBR (1).                                                                                                                                    ;
4.4.2.9.5              Statistical combination of Uncertainties (SCU)
Use of a 1.24 MDNBR limit with a best-estimate design CETOP model will. ensure, with at least 95% probability and 95% confidence,                                                                            p that the hot pin will not experience a departure from nucleate boiling. The 1.24 MDNBR limit includes explicit allowances for system parameter uncertainties, CHF correlation uncertainty, rod bow, the NRC penalties for the TORC code uncertainty and CHF correlation " prediction uncertainty," and a 0.01 penalty for the HID grids.
Several conservatisms are included in the SCU methodology (8),                                                                              B The significant conservatisms include                                                                                                            '
: 1. Combination                                      of    system    parameter                                  probability distribution functions at the 95% confidence level to yield a resultant MDNBR at >95% confidence.
: 2. Use            of                pessimistic            system  parameter                                  probability distribution functions.                                                                                                                ,
l Amendment F 1                                                                                  4.4-16                                    December 15, 1989 l
l l
 
CESSAR titMe n.n l
l 1
1
: 3.      Derivation of the new MDNBR limit such that it applies                                  i to both 4-pump and 3-pump operation.
: 4. Use of the single most adverse set of state parameters to generate the response surfacs.
: 5.      Application    of the CE-1 critical heat flux (CHF)                                      I correlation uncertainty based on the worst 16 x 16                                        )
assembly test section.                                                                    1
: 6. Application of the additional NRC CHF correlation                                        -
uncertainty penalty (" prediction uncertainty").
B 1
: 7.      Application      of    the  NRC-imposed              code  uncertainty penalty.                                                                                  !
: 8. Application of the 0.01 DNBR HID grid penalty imposed l'                  by NRC on the CE-1 CHF correlation.
4.4.2.10            Flur Tilt considerations                                                        Z t
An allowance for degradation in the power distribution in the X-Y                                      -
plane (commonly referred to as flux tilt) is provided in the                                          :
protection limit setpoints even though little, if any, tilt in the X-Y plane is expected.
The tilt,        along with other pertinent core parameters, is                                      "
!      continually monitored during operation by the COLSS (described in                                  i Section 7.7). If the core margins are not maintained, the COLSS actuates an alarm, requiring the operator to take corrective                                          '
action. The CPCs actuate a trip if limiting safety system                                          ,
settings are reached.
i The thermal margin calculations used in designing the reactor core are performed using the TORC and CETOP codes.                        The TORC and        B CETOP codes, which are described in Section 4.4.4.5.2, are based on an      open core        analytical    method for performing                  such calculations and treats the entire core on a three-dimensional                                        '
basis. Thus, any asymmetry or tilt in the power distribution is analyzed        by providing      the corresponding power distribution in the l B TORC  and CETOP      input.                                                                          "
4.
 
==4.3            DESCRIPTION==
OF THERKAL AND RYDRAULIC DESIGN OF THE REACTOR COOLANT SYSTEN (RCS)
A summary description of the RCS is riven in Section 5.1, Amendment B 4.4-17                          March 31, 1988
 
i CESSARti!Luo                                                                                                            l l
1 4.4.3.1        Plant configuration Data p
i An  isometric      view of the RCS is given in Figure 4.4-8.                                                          j Dimensions are shown on the general arrangement drawings, Figures                                                      !
5.1.3-1 and 5.1.3-2.              Table 4.4-6 lists the valves and pipe                                                '
fittings which form part of the RCS.                                                                                    i Table 4.4-7 lists the design minimum flow through each flow path in the RCS.                                                                                                            !
l Table 4.4-8 provides the volume, minimum flow area, flow path                                                          '
length, height and liquid level of each volume, and bottom elevation for each component within the RCS.
Components of the Safety Injection System (SIS) are located so as to meet the criteria for not positive suction head discussed in                                          B              :
i Section 6.3. Line lengths and sizes for the SIS are determined so as not to violate the fluid delivery rates assumed in the safety analyses described in Chapter 15.                          The total head losses throughout the injection lines are determined so as not to exceed                                                      l the head losses deduced from the fluid delivery rate.
Table 5.1.1-1 provides a steady-state pressure, temperature, and                                                        1 flow distribut:.on throughout the RCS.
4.4.3.2        oneratina Restrictions on Pumns The minimum RCS pressure at any given temperature is limited by the required net positive suction head (NPSH) for the reactor i
coolant pumps during portions of plant heatup and cooldown.                                        To                "
ensure that the pump NPSH requirements are met under all possible operating conditions, an operating curve is used which gives permissible RCS pressure as a function of temperature.
The reactor coolant pump NPSH restriction on this curve                                            is determined by using the NPSH requirement for ene pump operation (maximum flow, hence, maximum required NPSH) and correcting it for pressure and temperature instrument errors and pressure measurement location.              The NPSH required versus pump flow is supplied by the pump vendor. Plant operation below this curve is prohibited. At low reactor coolant temperature and pressure, other considerations require that the minimum pressure versus temperature curve be above the NPSH curve.
B Amendment F 4.4-18                        December 15, 1989
 
CESSAR8m%=.                                                                                                            ;
4.4.3.3        T_a=nerature-Power Operatine Man                                                                        ;
A temperature-power operating map (temperature control program) is provided in Section 5.4.10.
The adequacy of natural circulation for decay heat removal after                                                        -
reactor shutdown has been verified analytically and by tests on the Palisades reactor (Docket No. 50-255) and Calvert Cliffs Unit 1 (Docket No. 50-317). The core AT in the analysis has been shown to be lower than the normal full power ATr thus the thermal and mechanical loads on the core structure are less severe than                                                        '
normal design conditions.                In addition, St. Lucie Unit 1 (Docket                              y No. 50-335) and Palo Verde Unit 1 (Docket 50-528) successfully performed cooldowns from full power conditions using only natural circulation following reactor trip.
Heat removed from the core during natural circulation may be rejected by dumping steam to either the main condenser or the                                                      ,
atmospherer the rate of heat removal may be controlled to                                                          7'  ,
maintain core      T within allowable limits.                                                                      -  -
4.4.3.4          Load Followine characteristics The design features of the RCS influence its load following and                                                        '
transient response.        The RCS is capable of following the normal                                                  >
transients identified in Section 3.9.1.1. These requirements are considered when designing the pressurizer spray and heater                                                            ,
syatoms,    charging / letdown system, reactor regulating system                                                    1  ,
(RRS),    and    feedwater          regulating                system. Finally,                these transients are included in the equipment specification for each RCS component to ensure the structural integrity of the system.                                                        '
i When load changes are initiated, the RRS senses a change in the turbine power and positions CEAs to attain the programmed average coolant  temperature.        RCS          boron        concentration          can also be adjusted to attain the appropriate coolant temperature.                                                The feedwater system employs a controller which senses changes in steam flow, feedwater flow, and water level and acts to maintain steam generator level at the desired point.                                The pressurizer pressure and level control systems respond to deviations from preselected setpoints caused by the expansion or contraction of the reactor coolant and actuate the spray or heaters and the charging or letdown systems as necessary to maintain pressurizer pressure and level.
4.4.3.5        Thermal and Hydraulic characteristics Table Principal thermal and hydraulic characteristics                                of          the        RCS components are listed in Table 4.4-9.
Amendment F 4.4-19                            December 15, 1989 1_________ _____--_-_ _ ____ _ _    _  - - . - -          . . - . . -._-                          - _ - - . - .          -
 
CESSARtm%.m.                                                    -
1 I
4.4.4                  EVALUATION                                                                            I 4.4.4.1                  critical Beat Flux The margin to critical heat flux (CHF) or DNB is expressed in tarins of the DNBR.              The DNBR is defined as the ratio of the heat flux required to produce DNB at the calculated local coolant                                                !
conditions to the actual heat flux.                                                                          l The CE-1 correlation (1)(1) ine                was used with the TORC and CETOP                        g    ,
computer codes (1) to determ                      DNBR values for normal operation and anticipated operational occurrences.                            The CE-1 correlation was developed in conjunction with the TORC code specifically for DNB margin predictions for fuel assemblies with standard spacer l  grids              similar  to    those      in    System      80.      Topical    Reports CENPD-162(1) and CENPD-207(1) provide detailed information on the CE-1 correlation and source data, and comparisons with other data and correlations.                In brief, the correlation is based on data from tests conducted for C-E at the Chemical Engineering Research Laboratories of Columbia University.                                  Those      tests        used electrically-heated                5    x    5  array    rod    bundles    corresponding dimensionally to a portion of a 16 x 16 or 14 x 14 fuel assembly with standard spacer grids.                    The test programs conducted for the 16 x 16 and 14 x 14 geometries each included tests to determine the effects on DNB of the CEA guide tube, heated length, axial grid spacing, and lateral and axial power distributions.                                                    -
was developed from t
The DNB uniform data axial for power six CE-1 test correlation sections (1)ithw        the    following
* characteristics:
Heated        Axial Grid Fuel Assembly              No. Heated            Lateral Power        Length        Spacing Geometry          _ Rods              gistribution            (ft)          fin.)
16 x 16              25              Uniform              7                16.0 16 x 16              21              Nonuniform          7                18.3 16 x 16              21              Nonuniform          12.5            17.4 14 x 14              25              Uniform              7                14.3 14 x 14              21              Nonuniform          7                14.3 14 x 14              21              Nonuniform          12.5            14.3 l
I Amendment B 4.4-20                      March 31, 1988 l
 
CESSAR88% =
reactor fuel assembly (15 x 15) .      The validity of the inverse Peclet number of 0.0035 for the 16 x 16 assembly with standard grids was verified with data obtained in the tests conducted at Columbia University (1) .
The design basis requires that the minimum DNBR for normal operation and anticipated operational occurrences be chosen to provide a 95% probability at the 95% confidence level that DNB will not occur on a fuel rod having that minimum DNBR.
Statistical evaluation of the CE-1 correlation and relevant data shows that the appropriate minimum DNBR is 1.13 (1) (2) . Based on review of CENPD-162(1) and CENPD-207(1) , . the NRC requires use of    g a minimum DNBR of 1.19. Therefore, the minimum DNBR used for design is 1.19 for fuel with standard grids and 1.20 for fuel with HID grids. This limit was increased to 1.24 for System 80 as a result of the Statistical Combination of Uncertainties (SCU) lF analysis (see section 4.4.2.9.5).
4.4.4.2        Reacter Evdraulics 4.4.4.2.1        Reactor Flow Model Tests                              f The  hydraulic  design  of  the  System  80  reactor vessel    and internals . is supported by a three-phase flow test program with geometrically scaled models.      In the first phase, 1/8 scale air-flow model tests were conducted at Kraftwerk Union AG (KWU)          ,
to refine the geometry of the lower plenum and core support structure to attain an acceptable core inlet flow distribution.          s In these tests, geometric scaling was maintained up to the core inlet. The reactor core was represented by a single orifice plate matching the flow resistance through the lower end fitting and lower-most spacer grid, housed in a core shroud envelope.
The core inlet flow distribution was mapped by velocity probe measurements downstream of the orifice plate.        Because of the simplified core modeling and measurement technique, the KWU test results are considered to be preliminary.
In the second phase, 3/16 scale water-flow tests were conducted in  the C-E Nuclear Laboratories to refine the hydraulic performance of the upper plenum region, with respect to pressure drop and structural hydraulic loading. In these tests there was no representation of the reactor core.
In the third phase, a 3/16 scale water-flow model of the entire reactor and internals was tested to verify the design hydraulic parameters based on analysis and results of earlier tests.      This reactor flow model incorporates the minor design changes made after completion of the earlier model tests.      Model components are geometrically similar to reactor components, except for the core. Individual fuel assemblies are represented in the third Amendment F 4.4-23              December 15, 1989
 
CESSARim%ui..
test model by an array of square tubes. An axial distribution of orifice plates and of cross-flow holes in the double-wall boundaries between adjoining core tubes are sized to provide the                                                                  !
axial and lateral flow hydraulic resistance of the reactor core.                                                                -
This "open-core" flow modeling technique is a continuation of testing methods applied for the C-E 3410 MWt Series reactors (San                                                                -
Onofre Units 2 and 3, Forked River Unit 1, Waterford Unit 3, Pilgrim Unit 2), as described in CENPD-206(13) .                                    Details of the System 80 reactor flow model test and portions of the testa results are presented in Appendix 4A.                                                                                  lB Hydraulic design parameters derived from reactor flow model test                                                                '
results include:
o    The core inlet flow distribution and core-exit pressure distribution.                                                                                                ;
l o    Pressure drops in the reactor vessel.
1 o    Hydraulic loads on reactor internal components.
A.          Core Inlet Flow and Core Exit Pressure Distributions The core inlet flow and core exit pressure distributions are required as input to the TORC code for core thermal margin analysis (Refer to Section 4.4.4.5.2).                                                                              ,
The 4-loop core inlet flow distribution used in the TORC analysis is based in part on the results from the 1/8 scale lower plenum tests conducted at KWU. The flow distribution is characterized as having average or higher fuel assembly flow rates for the central assembly locations in the core,                                                          -
and lower than average assembly flow rates for the peripheral assembly locations.                  The core exit pressure distribution is based on an extrapolation of the pressure distribution measured in the 3410 MWt Class reactor flow model test program described in CENPD-206 (Reference 13).
The core exit pressura distribution is characterized as having lower than average exit pressures in the centrally located fuel assemblies and higher than average exit pressures in the peripheral fuel assemblies.
These core hydraulic boundary conditions were verified by the results from the 3/16 scale System 80 reactor flow model test. Flow model test information is used to define the core  inlet  flow distribution conditions for transients involving the shutdown of one or more loops.                                The test information was obtained from the 1/8 scale System 80 lower plenum model test and from model                              tests on earlier C-E reactor designs.
Amendment B 4.4-24                                      March 31, 1988
 
i.
CESSARUth l
l B.              Addition of the capability for handling non-zero lateral boundary conditions on the periphery of a collection of parallel flow channels.                        This capability is particularly important when analyzing a group of subchannels within the                                                          ,
hot fuel assembly.                                                                                                  '
C.              Addition of the capability to handle non-uniform core exit                                                          i
+                  pressure distributions.
O.              Insertion of standard C-E empirical correlations and the                                                          -
ASME fluid property relationsh;.ps.                                                                              -
Details of the lateral momentum equations and the empirical correlations used in the TORC code are given in CENPD-161 (j,) .
The application of the TORC code for detailed core thermal margin                                                                    4 calculations typically involves two or three stages.                                              The first B                    '
stage consists of calculating coolant conditions throughout the                                                                  !
core on a coarse mesh basis. The core is modelled such that the                                                                -:
smallest unit represented by a flow channel is a single fuel assembly.                  The three-dimensional power distribution in the core                                                  -
is superimposed on the core coolant inlet flow and temperature                                                                    '6 distr:.butions.
The core inlet flow anl core exit static pressure distributions are obtained from flow model tests discussed in Section 4.4.4.2,                                                                  '
and the inlet temperature                                  for normal four-loop operation is                                    .
assumed uniform.                  The axial distributions of flow and enthalpy in each fuel assembly are then calculated on the basis that the fuel                                                              u assemblies are hydraulically open to each other. Also determined during this stage are the transport quantities of mass, momentum                                                                  -
and energy which cross the lateral boundaries of each flow channel.
In the second stage, typically the hot assembly and adjoining fuel assemblies are modelled with a coarse mesh.                                                  The hot assembly is typically divided into four to five partial assembly regions.                One of these regions is centered on the subchannels rod adjacent                  to  the                        having  the    minimum        DNBR.                  The three-dimensional power distribution is superimposed on the core l  coolant inlet flow and temperature distributions.                                          The lateral transport of mass, momentum, and energy from the stage one calculations is imposed on the peripheral boundary enclosing the hot assembly and its neighbors. The axial distributions of flow and enthalpy in each channel are calculated as well as the transport quantities                    of mass, momentum, and energy which cross the lateral boundary                  of each flow channel.                  In some cases, the hot assembly detail                    normally included in the second stage is                                          B included in the first                    stage, thereby eliminating the need for the intermediate stage.                      In these cases, the second stage is the subchannel model discussed below.
Amendment B 4.4-33                    March 31, 1988
 
1 CESSAR titMemo.
l l
The  third    stage    involves a fine mesh modelling of the                              j partial-assembly region which centers on the subchannels adjacent to the rod having the minimum DNBR.                All of the flow channels used in this stage are hydraulically open to their neighbors.                            1 The output from the stage two calculations, in terms of the                                l lateral transport of mass, momentum, and energy is imposed on the lateral boundaries of the stage three partial assembly region.                            i Engineering factors are applied to the minimum DNBR rod and subchannel to account for uncertainties on the enthalpy rise and                          t heat flux due to manufacturing tolerances.                  The local coolant conditions are calculated for each flow channel.                        These coolant conditions are then input to the DNB correlation and the minimum value of DNBR in the core is determined.
A more detailed description of this procedure with example is contained in CENPD-161(1) . This procedure is used to analyze in detail    any  specific      three-dimensional power distribution superimposed on an explicit core inlet flow distribution. The                            '
detailed core thermal margin calculations are used primarily to develop and to support the simplified design core thermal margin calculational schema discussed below.                                                    ,
The method used for design calculations is discussed in detail in CENPD-206(13_). In summary, the method is to use one limiting hot
;                                                  assembly radial power distribution for all analyses, to raise or lower the hot assembly power to provide the proper maximum rod radial power factor, and to use the core average mass velocity in all fuel assemblies except the hot assembly.                        The appropriate reduction for the hot assembly mass velocity was determined by                          p the System 80 flow model tests (see Section 4.4.4.2.1).                          This methodology is used in the thermal margin analyses of the System 80 reactors.
The CETOP code (2), a variant of the TORC code, is used as a design code for System 80 thermal margin analyses. CETOP has the same theoretical bases as TORC, but has been improved to reduce B
execution time. The CETOP code uses the transport coefficients to obtain accurate determination of diversion crossflow and l
turbulent mixing between adjoining channels with a less detailed calculational model.        Furthermore, a predictor-corrector method                      i is used to solve the conservation equations, replacing the iterative method used in the TORC code,                  and thereby reduce                l execution time.      The conservatism of CETOP relative to TORC is                          l assured by benchmarking analyses which demonstrate that CETOP                              )
yields accurate or censervative DNBR results relative to TORC.                            1 l
l l
I
                                                                                                                                              )
l l
Amendment F          I 4.4-34                        December 15, 1909
                                                                                                                                              )
I
 
1 CESSAR tini?cer..                                                                                                        !
t TABLE 4.4 1
($heet1of2)                                                                    l THERMAL AND WYDRAULIC PARAMETERS Reactor Parameters                    _
System 80+ System 80 Waterford-3 i
Core Average Characteristics at Full Power:
Total core heat output, MWt                                                    3,800      3,800      3,390 Total core heat output, million Stu/h                                        12,970      12,970      11,570 Average fuel rod energy deposition                                            0.975      0.975      0.975 fraction Hot fuel rod energy deposition fraction                                        0.975      0.975      0.975 Primary system pressure, psia                                                  2,250      2,250      2,250          I  :
Reactor inlet coolant temperature, 'F                                            558        565          553            i Reactor outlet coolant temperature, 'F                                          615        621          611 Core exit average coolant temperature, 'T                                        617        624          613        -
Average core enthalpy rise, Btu /lbm                                              81          82            81      -  i w  .
Design minimum primary coolant flow                                        445,600      445,600    396,000 rate, gpm.
Design maximum core bypass flow, % of                                            3.0        3.0          3.5 primary Design minimum core flow rate, gpm                                        432,200      432,200    382,000 Hydraulic diameter of nominal subchannel,                                      0.471      0.471      0.471 in.                                                                                                          p Core flow area, ft 2                                                            60.8        60.8          54.7 Coreavgmgssvelocity,million                                                    2.64        2.62          2.61 lbm/h-ft Core avg coolant velocity, ft/s                                                16.6        16.7          16.3 Core avg fuel rod heat flux, Btu /h ft 2                                  185,100      185,1008 )  182,400 Total heat transfer area, ft 2
68,320          68,320a )  62,000 a) Corrected values for System 80 design Amendment F December 15, 1989
 
  .i CESSARMRLm.m TABLE 4.4-1 (Cont'd)
(Sheet 2 of 2)
THERMAL AMD.. HYDRAULIC _ PARAMETERS Reactor Parameters              _
System 80+ Lyi.ttg_Ag Waterford-3 Average fuel rod linear heat rate kW/ft                    5.42      5.42        5.34 F
Pcuerdensity,kW/ liter                                  95.5        95.5        94.9 No. of active fuel rods                                54,764      54,764      49,580 Power Distribution Factors:
Rod radial power factor                                    1.55      1.55        1.55 B
Nuclear power factor                                      2.28      2.28        2.28 Total heat flux factor                                    2.35      2.35        2.35 Maximum augmentation factor                            1.059      1.059        1.041 Maximum gap length, in.                                0.761      0.761        1.20 Engineering Factors:
Engineering heat flux factor                              1.03      1.03        1.03 Engineering enthalpy rise factor                          1.03      1.03        1.03 Pitch, Bowing, and Clad Diameter Enthalpy                  1.05      1.05        1.05 Rise Engineering factor on linear heat rate                    1.03      1.03        1.03 Charteteristics of Rod and Channel with Minimum DNBR:
Maximum fuel rod heat- flux, Btu /h-ft2              434,900    434,900)  a 428,000 Maximum fuel rod linear heat rate, kW/ft                  12.7      12.7        12.5  F 00 maximum steady state temperature, 'F                3,200      3,205")    3,180 2
Outlet temperature, 'F                                  640.1      645.78 )      642 lB a              F Outlet enthalpy, Btu /lbm                              676.6      687.l )        680 Minimum DNBR at nominal conditions                        2.10      1.988 )      2.07 (CE-1 correlation)                                                                        B a) Based on updated System 80 flow distribution Amendment F December 15, 1989
 
CESSAR FRP#un.                                                                                                ,
f TABLE 4.4-2                                                                ,
COMPAR;: SON OF TH: DEPARTJRE FRON NUC.EATE ROILING RAT '05 COMPUTI :D WITH 11FFERENT C0' TRELAT 10NS DNBRs for Reactor                        i Conditions Giving a                        ,
1.24 CE-1
* DNBRs for Nominal                          Minimum DNBR in Reactor Co9ditiens                        Matrix Subchannel Subchannel                                subchsnnel                +
Natrix                  Next to          Natrix                Next to              !
Correlation          Subchannel        Guide Tube              Subchannel        Guide Tube      p CE-1                        2.26                    2.10              1.24                  1.12 Original W-3(4)              2.60                    2.80              1.20                  1.26 l        RevisedW3(1)                2.60                    2.28              1.20                  1.19 l
i                                                                                                                          ,
Amendment F December 15, 1989
 
    ,, i o
CESSAR !Nitricui                                                                                                        j l
1 i
j TABLE 4.4-4                                                              ]
REACTOR VE55EL BEST ISTIMATE tRE$11dtLLO55E5 AtgLIJt01&hLIDERATE1                  J                                      j Pressuregoss                      Tamparature Connonent                                  (1b/in. i                          f'F)                    >
Inlet nozzle and 90' turn                                          7.7                              558 Downcomer, lower plenum, and support i        structure                                                        16.6                              558 F            '
Fuel assen.bly                                                    16.8                              589 t
Fuel assembly outlet to outlet norile                                                            16.J1                              617 Total pressure loss                                    57.9 Amendment F
!                                                                                                December 15, 1989
 
t -
l CESSAR ttWien...
I I
J l
TABLE 4.4-6                                            1 4
RCS VALVES AND PIPE FITTINGS                                  i
                                        . Pressure Boundary Valves -
Valve                      Valve No.            $1re (in.)      Quantity Pressurizer Safety Valves              RC 200, RC-201              6X8              4 RC-202, RC-203            ,                        ;
Pressurizer Spray Control Valves      RC-100E, RC-100F                3            2 Spray By,rass Needle Valves            RC 236, RC 237                3/4            2 Refueling Level Indicator              RC 214                        3/4            1 l
Connection Isolation Valve I
Reactor Vessel Head Vent Isolation    RC 212                        3/4            1 Valve                                                                                    j All other RCS valves are identified in Section 5.4. The Safety Depressurization        F
!          System valves are identified in Section 6.7.
i
                                            - RCS Pipe Fittings -                                    ,
Elbows                Size fin.)                    Radius fin.)              ouantity 35'                      -42                          63                          2 1          45'                        30                          45                          4 90'                        30                          45                          8 44'9'                      30                          45                          4 Amendment F December 15, 1989
 
.,u, ,. .      .. . . . . . .
                                                                                                                              '    ~    '
                                                                                                                                                        ^
a TABLE 4.4-8 REACT 0g cans _4NT SYSTEN GE0fETRY Top Elevation          Bottom Elevation      Hintanas Flow -          Volemma-Flow Path                                                                  2-                3 Lenoth fft)        -(d1 ~    -fft)        fdi          ffti      Area fft 3              fft 1
                              , monent 2.38                  - 1.75
                                                                                                          ~
9.62                  135.27' Hot Leg                                  14.06                                                              4.91                  '119.38.
24.32                        0.58                  - 9.97 Suction Leg                                                                                  - 1.25        4.91                    94.74-Discharge Leg                            19.30                        1.25 2400 Pressurizer                                                            (e)                                      E 50.07 *)            -1200'    r.,
Liquid. Level (full power)                                          (e).                      (e) 1.75-    0.56                    43.62 Surge Line                              77.91                          (e)
Steam Generator                                                        3.90                  - 0.48        9.62                    25.55 Inlet Nozzle                            3.07                                                            4.91                    12.67 2.79                        2.41                  - 1.19 Outlet Nozzle                                                      6.48                  - 0.10      19.07                  410.98 8T Inlet Plenum                            4.74(b)                                                          9.74                  410.98 4.74(b)                    6.48                  - 0.10' Outlet Plenum 61.15                      40.94                        6.48      0.002ICI            1939.60 Tubes (Active & Inactive)
Reactor Vessel                                                        1.4-                  - 1.5        4.9                    21.7 Inlet Nozzle (ea)                      3.7                                                            33.8                  1157.1 21.4                        11.7                    -22.6 Downcomer
                                                                              -20.5                    -25.9        32.5                    430.2 Lower Plenum                            3.2                                                            44.4                    239.2 2.8                      -17.7                    -20.5 Lower Support Structure &
Inactive Core                                            - 5.3                    -17.8        60.8                    888.2-Active Core                            12.5                                                            46.3                    262.9 2.8                      - 2.5                    - 5.3 Upper Inactive Core                                                2.1                    - 2.4        26.6                    459.4 Outlet Plenum                          5.7                                                              0.1                  .240.6 15.9                      - 2.7                    -19.6 Core Shroud Bypass                                                                        - 3.5        0.4                  1352.5-CEA Shroud Assembly & Tie              17.9                        15.6 Tubes 12.7                        2.1      1.6                    226.0 UGS, CEA Shroud Annulus                10.6                                                              7.8                    422.6 3.2                        19.9                      12.7 Top Head
                                                                                '1.7                    - 1.8        9.6                    32.2-Outlet Nozzle                          4.0 For the cylinder.                                          d. Reactor Vessel nozzle centerline is the reference
: a.                                                                          elevation. It has an elevation of 0.0 ft.                      F.
: b.      Represents a geometrical rather than an                            See Section 5.4.
actual flow path length.                                    e.
: c.      Flow path area per tube.                                                                          Amendment F-December 15, 1989
 
c                                ,
p fil 1CESSARBRL m h      .e.
f
    ,                                                      TABLE 4.4-9' (Sheet 1of2)-                                  .
      ,                                        ltEACTOR C0OLANT SYSTEM COMP 0NENT THERMAL AND HYDRAULIC DATA
        -                        *Connonent                                            Data Reactor ~ Vessel' Rated core thermal power, MWt                                  3,800-
: Design pressure, psia                                          2,500 Operating pressure, psia-                                      2,250 Coolant outlet temperature, 'F                                615                B Coolant inlet. temperature, 'F                                558 Subcooled
                . Total.
Coolant    outlet flow,10 coolant  state '6 lb/h 165.6
                ' Average- coolant enthalpy                                                        i
                    ~ Inlet, Btu /lb                                            557                B l0utlet, Btu /lb                                            635 Averagecoola$ density Inlet, lb/ft 3                                            46.3 Outlet, lb/ft -                                            41.8:
Steam Generators Number.of units                                                2 Primary. Side-(or tube side)
Design pressure / temperature, psia /*F                    2,500/650 Operating pressure, psia                                    2,250
                      -Inlet temperature, 'F                                      615 Outlet temperature, 'F-                                    558 Secondary Side (or shell side)
                    -Design pressure / temperature, psia /'F                      1,200/570          B Full load steam pressure / temperature, psia /*F            1,000/545
                    ~ Zero load sterm pressure, psia                              1,100 6
Total steam ' sow per gen.,1b/h                            8.56 x 10 Full load steam quality, % (minimum)                        99.75              F Feedwater temperature, full power, 'F                      450 Pressurizer Design pressure,-psia                                          2,500 Design temperature, 'F                                        700 Operating pressure, psia                                      2,500 Operatingtemperatuge,'F                                        647                B Internal volume (ft )                                          2,400 Heaters Type and rating of heaters, kW                              Immersion /50 Installed heater capacity, kW                              1,800 Amendment F December 15, 1989
 
3 LCESSAR m a mm TABLE 4.4-9(Cont'd)
(Sheet 2of2)
REACTOR COOLANT SYSTEM COMP 0NENT THERMAL AND HYDRAULIC DATA Component                                          Data Reactor Coolant Pumps Number of units                                          4 Type                                                      Vert.-Centrifigual Designcapacity,(gpm)                                      111,400 Design pressure / temperature, psia /*F                  2,500/650 Operating pressure, psia                                  2,250-T,,pe drive                                              Squirrel cage induction motor
      -, ;    Total dynamic head, ft                                    365 Rating and power requirements, hp, hot                    8,850 Pump speed, rpm                                          1,190 Total heat input to RCS, MWt                              17 Reactor Coolant Piping 6
Flow per loop (10    lb/h)
Hot leg                                                82 Cold leg                                              41 Pipe size (inside dia.), in.
Hot leg                                                42 Cold leg Suction leg                                          30
          -        Discharge leg                                        30 Pipe design press./ temp., psia /*F                        2,500/650 Pipe operating press./ temp., psia /*F Hot leg                                                2,250/615        B Cold leg                                                2,250/558 Amendment B March 31, 1988
 
7;7                                              7Q
                                                                                        /  ,,
b                            -
                                                                                        /u                                        i      a
                                                                                                                                                    +
                                                                                          -l              _                _/ k              M
                                                                                              -5    3      .
t                                \
                                                                                                            !": 1)
S@                                                                :
N
                                                                            \
bl.7 lR
                                                                                                                                                                !s l
i s                    i            .
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(                            M                                                                '
L j                        L                Y @*                                      k
                          ' TOTAL FLOW                                    hl                                                            ll Il l  ll                    8                                            '
    .                                                                k      N N  :
                                                                                  .ams:
                                                                                  )
l
                                                                                                                                  "!!E7                  l s
N
                                                                            \                                                                                    \
                                                                            \                                                                                l  \
l                                                                      1
                                                                            \                                                                                l  s
                                                                            \                                                                                j \
                                                                                  '                                                                              \
k                                                                                                h MAIN FLOW                                    ,
                                                                            \                                                                            (        N          ~ BYPASS FLOW
                                                                                  )
                                                                            \                                                                                    \
i  I
                                                                                                                        , d h,                                    s 1 THROUGH OUTLET NOZZLE CLEARANCE s#
                                                                                                            @@                                                    \
2 THROUGH INSTRUMENTED CENTER s  N          -                            <      :
                                                                                                                                                                  \
GUIDE TUBES N'li
                                                                            \J 8      FFF                :
I Q)- **                  ;
                                                                                    =
                                                                                      -          ,1 0, ~,,:    ,[                                  '.                3 THROUGH SHROUD BARREL ANNULUS 4 THROUGH CENTER GUIDE TUBES
                                                                                \
g"k./        "      "    '    d                      #./                    5 THROUGH OUTER GUIDE TUSES N                                  I
                                                                                                        ) !          ,
6 THROUGH ALIGNMENT KEY. WAYS t      s
                                                                                          ;            4        1i                  i l
Amendment F                i Drcomber 15,1989 Figure          I REACTOR FLOW PATHS                                        4,4 6
      . _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ .                                                                                      ._      _ _        __ _-.. . ..- _ .          .-    .-.._._..-.\
 
                                                                                  }
m  fCESSAR Emincim,.
1 l
4.5          KEACTOR MATERIALS 4.5.1          CONTROL ELEMENT DRIVE STRUCTURAL MATERIALS                  ,
1 4.5.1.1          Material snecifications A.      The materials. used in the control element drive mechanism        i
                '(CEDM) reactor coolant pressure boundary components are as follows:
                -1. Motor housing assembly SA 182, Type 347 (austenitic stainless steel)                  !
ASME Code Case N-4-11 (modified Type 403 martensitic F
stainless steel), and additional requirements of ASME SA-182 SB 166 (nickel-chromium alloy)
: 2. Upper pressure housing SA 213, Type 316 (austenitic stainless steel)
SA.479, Type 316 (austenitic stainless steel)            l0 The above listed materials are also listed in Section III of the ASME Boiler and Pressure Vessel Code. In addition, the      p materials comply with Sections II and IX of the ASME Boiler and Pressure Vessel Code. Code Case N-4-11 is acceptable per Regulatory Guide 1.85.
l                The functions of the above listed components are described in Section 3.9.4.1.                                              ,
B.      The materials in contact with the reactor coolant used in the CEDM motor assembly components are as'follows:
: 1. Latch guide tubes ASTM A269, Type 316 (austenitic stainless steel)
Chrome oxide (plasma spray treatment)
: 2. Magnet and spacer ASTM A276, Type 410 (martensitic stainless steel)
Amendment F 4.5-1              December 15, 1989
 
                  'CESSAR !!iWrie.m.,
i
: 3.      ' Latch:and magnet housing ASTM A276, Type 316 (austenitic stainless steel) l QQ-C-320, Class 2B (chrome plating)                                          D-ASTM A276, Type 440C (martensitic stainless steel)-                                        [
i
: 4.      -Spacer l                                      ASTM A240, Type 304 (austenitic stainless steel)
: 5.        Alignment Tab ASTM A276, Type 410 (martensitic stainless steel)
: 6.          Spring AMS 5698B, Inconel X-750 (nickel base. alloy)
: 7.          Pin Haynes~ Stellite No. 6B-(cobalt base alloy)=                          .
: 8.          Dowel pin ASTM A314, Type'413 (martensitic stainless. steel)
          .                9.          Spacer and screw l                                      ASTM A276, Type 321 (austenitic stainless steel)
: 10.        Stop ASTM A276, Type 304 (austenitic stainless steel) l                            11.        latch and pin Haynes Stellite No. 36 (cobalt base alloy)
: 12.        Locking cup and screws Type 300 Series austenitic stainless steel
: 13.        Steel Ball F
ASTM Ak76, Type 440C The functions of the CEDM motor assembly components are described in Section 3.9.4.1.
Amendment F 4.5-2                          December 15, 1989
  -_:___. _ _ _ - _ _ _ _ _ = _ _ _ _ _ _ _ _ _    -    . - - - . - . . .    . - - - . .              .  ..
 
m wlCESSARUN%m<
[
Weld: heat affected zoneL sensitized austenitic stainless steel'
                                                    ~
(which will fall in the Strauss Test,: ASTM A708) - is avoided in control elementidrive mechanism structural components by careful-control of:
                                      ~
A ~. - Wald heat input to less than 60 kJ/in B. Interpass-temperature to 350'F maximum C. Carbon content to s 0.065                                          0 4.5.1.4          control of Delta Ferrite in Austenitic Stainless Steel Welds The austenitic stainless steel,-primary pressure retaining welds 4          - in the control element drive' mechanism structural components are            j consistent with the recommendations of Regulatory Guide 1.31 as follows:
The delta ferrite content' of A-No.8 (Table 2W-442 of the ASME Code, Section IX) austenitic stainless steel welding materials is controlled to 5FN-20 Fit.-                                                D The delta ferrite determination is carried out using methods.                j' t            specified.in the ASME Cocli, Section III, for each heat, lot or L
heat / lot combination af welJ filler material..      For the submerged      =i arc process, the delta = ferrite determination for each wire / flux            .
combination    may: be    ande    on  a  production  or  simulated (qualification) production weld.
4.5.1.5        cleanine and contamination Protection Procedures The    procedure  and  practices      followed- for    cleaning- and
            - contamination protection of the control element drive mechanism
            . structural components are in compliance.with the recommendations of Regulatory Guide 1.37 (including ANSI /ASME NQA-2-1983) and are described below:                                                          F
            - Specific    requirements      for  cleanliness    and    contamination        i protection are included in the equipment specifications for components fabricated with austenitic stainless steel.              The        '.
provisions described below indicate the type of procedures
: l.          utilized for components to provide contamination control during
            - fabrication, shipment, and storage.                                          i i
Contamination of austenitic stainless Lteels of the Type 300 series by compounds that can alter the physical or metallurgical structure and/or properties of the material is avoided during all stages of fabrication.        Painting of Type 300 series stainless steels is prohibited.        Grinding is accomplished with resin or Amendment F 4.5-5                December 15, 1989
 
CESSARlHMeim
                                                                                      'l rubber-bonded aluminum oxide or silicon carbide wheels that have not previously been used on materials other than Type 300 series                    I stainless steel alloys.                                                            l Internal surfaces of- completed components are cleaned to the extent that grit, scale, corrosion products, grease, oil, wax gum, adhered or embedded dirt, or extraneous material are not                      l visible to the unaided eye.                                                        !
Cleaning is effected by either solvents - (acetone or isopropyl                    ,
alcohol) or inhibited water (300-200 ppa hydrazine). Water will                    !
conform to the following requirements:                                            !
l
* Halides Chloride, ppm              < 0.60 Fluoride, ppm              < 0.40 Conductivity, pmhos/cm            < 5.0 l
pH                              6.0 - 8.0 Visual clarity                  No turbidity, oil or sediment To prevent halide-induced integranular corrosion that could occur in  an aqueous environment with significant quantities of                        l dissolved oxygen, flushing water is inhibited via additions of                    ;
hydrazine. Experiments have proven these inhibitors to be                      I affective. Operational chemistry specifications preclude halides                    l and oxygen (both prerequisites for intergranular attacks) and are                  I shown in Section 9.3.4.
4.5.2        REACTOR INTERNALS MATERIALS                                          l l
4.5.2.1        Material snacifications
    - For reactor internals, the material specifications satisfy the requirements of Article NG-2000 in Section III of the ASME Code.            F The materials used in fabrication of the reactor internal i    structures are primarily Type 304 stainless steel.                The flow l    skirt is fabricated from Inconel.          Welded connections are used I
where    feasible;    however,    in    ' locations    where      mechanical connections are required, structural fasteners are used which are designed to remain captured in the event of a single failure.
Structural    fastener material      is    typically a high        strength        l austenitic    stainless    steel;    however,    in    less      critical        I upplications Type 316 stainless steel is employed.            Hardfacing of
: l. Stellite material is used at wear points.                The effect of r    irradiation on the properties of the materials is considered in l  the design of the reactor internal structures.            Work hardening            I properties of austenitic stainless steels are not used.
Amendment F                .
4.5-6                    Decerber 15, 1989        l L
 
  .w 7
CESSAR!!BL mw The following is- a list of the major components of the reactor 4    -internals together with their material specifications:
A. Core support barrel assembly-
: 1. Type- 304 ' austenitic stainless steel to the following specifications:-
: a. ASTM-A-182
: b. ASTM-A-213
: c. ASTM-A-240
: d. ASTM-A-479
: 2. Precipitation hardened stainless steel to_the following specifications:
: a. ASTM-A-453, Grade 660
: b. ASTM-A-638, Grade 660 B.. Upper guide structure assembly
: 1. Type 304 austenitic stainless steel to the following
                -specifications:                                              -
: a. ASTM-A-182-
: b. ASTM-A-240-
: c. ASTM-A-213
: d. ASTM-A-479                                          D t
: 2. Precipitation hardened stainless steel to the following      +
specification:
: a. ASTM-A-638, Grade 660 C. Core shroud assembly
: 1. Type 304 austenitic stainless steel to the following specifications:
: a. ASTM-A-182
: b. ASTM-A-240 D. Holddown ring ASTM-A-182, modified to ASME Code Case N-124. This alloy is D
tempered at 1040 to 1120*F in accordance with Code Case N-124 requirements.
Amendment D 4.5-7              September 30, 1988
 
        +      '
CESSAR Ennneamm E.;  - Bolt and pin material                                                ;
ASTM-A-453 and ASTM-A-638, Grade 660 material (trade name A-286) is used for bolting and' pin applications. . Thin' alloy        I is heat treated-in-accordance with the ASTM specifications        D  j
                  -by precipitation hardening at 1300-1400*F for 16 hours to a            I minimum yield strength of 85,000 psi.            Its corrosion        l properties are similar to - those of the Type 300 series            1 austenitic stainless steels.      It is- austenitic in all            !
conditions of fabrication and heat treatment.        This - alloy      I was used for bolting -in previous reactor systems and test            I facilities in contact with primary coolant and has proven              '
completely satisfactory.                                              ;
1 F. Chrome plating and hardfacing                                          '
Chrome plating -or hardfacing are employed on reactor                '
internals components or portions thereof where required by            )
function.      Chrome  plating    complies    with    Federal Specification    No. QQ-C-320. The  hardfacing employed is Stellite 25.                                material  l0 All    o
                  ~f  the materials employed in the reactor internals' and in-core instrument support system have performed satisfactorily in operating reactors such as Palisades (Docket-50-255), Fort
          .Calhoun (Docket-50-285) and Maine Yankee (Docket-50-309).
L L          4.5.2.2          Weldina Accentance Standards Welds employed on reactor internals and core support structures                l are fabricated in accordance with Article NG-4000 in Section III,        p  !
and meet the acceptance standards delineated in.artic3e-NG-5000,              I Section III,. Division I, and control of welding is performed in              !
I          .accordance with Section III, Division I, and Section IX of the                )
l-        ASME. Code. In addition, consistency with the recommendations of Regulatory Guides 1.31 and 1.44 is' described in Section 4.5.2.3.
l 4.5.2.3          Fabrication and Processine of Austenitic Stainless Steel The following information applies to unstabilized austenitic stainless steel as used in the reactor internals.
4.5.2.3.1          Control of the Use of Sensitized Austenitic Stainless Steel The recommendations of Regulatory Guide 1.44, as described in Sections 4.5.2.3.1.1 through 4.5.2.3.2.5, are followed except for the - criterion used to demonstrate freedom from sensitization.
,    ,    The ASTM A708 Strauss Test is used in lieu of the ASTM A262 l          Method E, Modified Strauss Test, to demonstrate freedom from Amendment F 4.5-8                December 15, 1989
 
CESSAR ?!nincan.                                                                                          q i
l 4.6        FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 0
The System 80+ Standard Design includes the following reactivity control systems: the control element drive mechanisms (CEDMs),
the safety injection system (SIS), and the chemical and volume control system (CVCS). The CEDMs are referred to collectively as the control rod drive system (CRDS) . The pertinent information, evaluations, 'and testing of the CRDS are treated in Section 4.6.1,  4.6.2,    and 4 . 6. 3 respectively.      The combined performance                                '
of the CRDS and other reactivity control system is discussed in Sections 4.6.4 and 4.6.5.
4.6.1        INFORMATION FOR THE CONTROL ROD DRIVE SYSTEM (CRDS)
The CRDS consists of the CEDMs.            Component diagrams, description, and characteristics of_the CEDMs are presented in Section 3.9.4.
4.6.2        EVALUATION OF THE CRDS                                                            9 2
The safety function of the CRDS is to drop CEAs into the reactor core when the motive power is removed from the CEDM power bus.
The active interface between the RPS and the CRDS is at the trip circuit breakers located in the reactor trip switchgear (RTSG).                                          o 4.6.2.1          Sincie Failure A failure mode and effects analysis of the RPS (including the l    RTSG) is presented in Section 7.2, which demonstrates compliance j    with IEEE Standard 279-1971, and shows that no single failure in o    the RPS can prevent the removal of electrical motive power from
!    the CEDMs.      For the trip function, the CEDMs _ are essentially l    passive devices.        When power is removed from the CEDM coils, the l    armature springs automatically cause the latches to be disengaged l  from the CEDM drive shafts, allowing insertion of the CEAs by l    gravity.  -For the execution of the trip function, all the CEDMs
: l. are independent.of one another.            In other words, the failure of one CEDM to trip does not affect the operability of any other CEDM. Sufficient shutdown margin is always maintained to assure that the shutdown capability can be retained in the event of a failure of any CEDM.        Therefore, no single failurs can prevent the CRDS from providing sufficient scram reactivity to achieve a shutdown.
4.6.2.2          Isolation of the CRDS from other EcuiDment The interface between the CEDMs and the CEDM Control System is at the CEDMCS power switches, which provide the isolation of the motive power from the low voltage logic control signal.                                  The interface    between      the      CEDMs  and  the    CEAs      involves              no non-essential elements.          Therefore, no isolation is required.
l                                                                  Amendment B 4.6-1                  March 31, 1988
 
L"    CkSSARtmiPhi.&
4.6.2.3      -Protection from Common Mode Pailure 4.6.2.3.1        Pipe Breaks-Protection. of essential systems from' the consequences ~ of. a        i postulated pipe rupture is described in section 3.6.-
F-l 1
                                                                              'l l
l, I
p m
i l
Amendment F 4.6-2            December 15, 1989 1
1
 
I LCESSAR tinineau.=
        \
          '4.6.3        TESTING AND VERIFICATION OF THE CRDS                                                                    ,
The pre-core and post-core-CEDM performance test is described in Chapter 14, which verifies the proper operation and sequencing of                                                    4 the CEDMs.
4 . 6 .' 4    INFORMATION FOR COMBINED FERFORMANCE OF THE REACTIVITY CONTROL SYSTEMS
_ Plan . and elevation layout drawings of the reactivity _ control systems are presented in the site-specific SAR.                                                                      1 Table 4.6-1 lists all the postulated -accidents analyzed in
!          Chapter 15 that take credit for two or more reactivity control-l'        systems for preventing or mitigating each accident.                                              The related reactivity systems are also tabulated.
4.6.5'        EVALUATION OF COMBINED FERFORMANCE The CRDS and SIS are separated and totally diverse in design and
,          operation. - In addition, since..the_CRDS and the SIS are protected
!          from missiles, pipe breaks and their effects, (as delineated in Section;6.3) there are no credible potential common-mode failures that could cause the combination of the CRDS and. SIS to fail to.                                              p' provide sufficient reactivity insertion to achieve a reactor l
shutdown.
As . described , in Section 9.3.4.1,                    the CVCS is a non-safety-grade system and is not required to perform any accident mitigation or l        safe shutdown function.        The CVCS is, however, designed for a L        high degree of redundancy and reliability.
L B
Amendment F 4.6-3                                                      December 15, 1989 l
 
f .-u g                                                              .  .
                                                            . (Shost l'of 6)
      ~
(        CESSAR %IELuo EFFECTIVE PAGE LISTING APPENDIX A Table of contents.
gggg,                                          Amendment Overview _-                                          F i'                                                  F F
ii-1911 gggg                                            Amendment A-1                                                  F A-2                                                  F A-3_                                                F-A                                                F                    s --
A
* F
        'A-6.                                                F A-7                                                  F A                                                F A-9                                                  F                    ,.
A-10                                                F                    i A-11                                                F A-12                                                F                    ,
A-13                                                F                  - 4.
A-14                                                F A-15                                                F
        'A-16                                                  F
        .. A-17                                                F A-18                                                F A-19                                                F A-20                                                F A-21                                                F A-22                                                F A-23                                                F A-24                                                F A-25                                                F
          -A-26                                                F A-27.                                                F A-23                                                F A-29                                                  F A-30                                                F A-31                                                F A-32                                                F A-33                                                F i
Amendment F December 15, 1989
 
:            LCESSAR tiEnemt.                                                                                                                          .,
t EFFECTIVE PAGE LISTING -(Cont'd)
APPENDII A                                                                                          ,
Zag.t (Cont'd) r                                                                                                                                                ;
ZAgt                                                                                    Amendment                                          ,
A-34                                                                                            F-
* A-35                                                                                            F
                  ~A                                                                                            F A-37                                                                                            F                                        ,
A-38                                                                                            F                                        "
A-39                                                                                            F A-40                                                                                            .F A-41                                                                                            F
                  .A-42                                                                                            F-A-43~                                                                                            F
                  -A                                                                                            F
                  -A-45                                                                                            F l
A 46                                                                                            F
                  ~A-47      *
                                                                                                                  .F A                                                                                            F
                  .A                                                                                            F                                        ,
A-50                                                                                            F A                                                                                            F
  ;                A-52'                                                                                            F-A-53                                                                                            F A-54                                                                                            F A-55                                                                                            F                                        ,
i:                A-56                                                                                            F 4                  A-57                                                                                            F L                =A                                                                                            F                                        "
l                  A-59                                                                                            F A                                                                                            F l
L                  A-61                                                                                            F l                  A-62                                                                                            F.
l                  A-63                                                                                            F l                  A-64                                                                                            F
                  -A-65                                                                                            F A-66                                                                                            F
                  'A-67                                                                                            F
                  .A-68'                                                                                            F A-69                                                                                            F A-70                                                                                            F                                        l
                  'A-71                                                                                            F A-72                                                                                            F A-73                                                                                            F A-74                                                                                            F Amendment F December 15, 1989
 
i ,
                    >g                                                                                      ;
t                                                                                      .
h                                                                          (ShGot 3'of'6).
CESSAR !!Ihn.(                                                                  .,
1.i .
l 1
EFFECTIVE PAGE LISTING (Cont'd)
APPENDIE A 2331 (cont'd)
P Page                                            Amendment A-75                                                  F A-76                                                  F A-77'                                                F                        3 A-78                                                  F A-79                                                  F
                      .A-80                                                  F
                        .A-81                                                  F A-82                                                  F A-83                                                  F L                        A-84                                                  F
                      ~A-85                                                  F                          i
                      .A-86                                                  F A-87                                                  F                ,
:. A-88 '
F A
* F                          ,
A-90                                                  F' A-91                                                  F                          '
i-                      A-92'                                                F l'                      A-93                                                  F L                        A-94                                                  F
: l.                      A-95                                                  F A                                                F A-97                                                  F l-                      A-98                                                  F                          ,
L                        A-99                                                  F
: l.                    <A-100                                                  F A-101                                                F A-102                                                F A-103                                                F A-104                                                F A-105                                                F A-106                                                F A-107                                                F A-108                                                F A-109'                                                F l
                      'A-110                                                  F A-111                                                F i-                      A-112                                                F l'                      A-113                                                F A-114                                                F A-11S                                                F Amendment F December 15, 1989 V    *'ee -
ww* e=,-ew-    -+a- - -                            -
 
_ _                      _(
                      #                                                                                                                      )
                                                                            .(Shoot'4.of 6)
CESSAR Mne.mm                                                                                                                        J h
    ;                                                                                                                                        v l.i EFFECTIVE PAGE LISTING (Cont'd).
y..                                                  APPENDII A.-
Test , (Cont'd)                                                                                                              ,
Rage                                                          Amendment
* A-116                                                              F-A-117-                                                            F
        - A-118-          -
F A-119:                                                            F A-120                                                              F A-121                                                              F                                                              ,
i; A-122                                                              F A-123-                                                            F A-124                                                              F A-125-                                                            F A-126                                                              F A-127-                                                            F-L        A-128                                                              F L        A-129                                                              F 1
A-130                                                              F
        ' A-131 -                                                            F                                                              -
A-132                                                              F A-133                                                              F A-134                                                              F
        ' A-135                                                            F
('      ' A-136                                                              F                                                              ,
F A-137 A-138                                                              F A-139                                                              F
        ' A-140                                                              F l      . A-141-                                                            F l'        A-142                                                              F A-143                                                              F
!        .A-144                                                              F A-145                                                              F L
A-146                                                              F A-147                                                              F A-148                                                              F
        ' A-149                                                              F
        ' A-150                                                              F A-151                                                              F A-152                                                              F E        A-153                                                              F A-154                                                              F l'        A-155                                                              F
!        A-156                                                              F i
Amendment F t.
December 15, 1989
 
      >x    .
i i.-
LC E S S A R E ncuio.                                                                  1 ,'
I i
c EFFECTIVE PAGE' LISTING (Cont'd)                                    ;
t=
APPENDIX A TAKt (Cont'd)
Raga                                                    Amendment A-157                                                        F A-158                                                        F Tables                                          Amendment Al-1 (Sheet 1)--                                            F Al-1 (Sheet 2)                                              F Al-1 (Sheet 3)                                              F Al-1.(Sheet 4)J                                              F l              Al-1 (Sheet 5)                                              F Al-1 (Sheet 6)                                              F Al-1 (Sheet 7)                                              F Al-1 (Sheet 8)-                                              F Al-1 (Sheet'9)                                              F.
F
              'Al-1 (Sheet 10)                                              F'
              .Al-1;(Sheet'11)
Al-1f(Sheet'.12)                                            F-Al-1 (Sheet 13)                                              F
              'Al-1 (Sheet 14)                                              F Al-1 (Sheet 15)                                              F Al-1.(Sheet'16)                                              F                        '
                -Al-1 (Sheet 17)                                            F Al-1 (Sheet 18)                                            F Al-1 (Sheet 19)                                            F Al-1 (Sheet 20)                                            F Al-1 (Sheet 21)                                            F Al-1 (Sheet 22)                                            F Al-1'(Sheet 23)                                            F Al-1 (Sheet 24)                                            F Al-1 (Sheet 25)                                            F Al-1 (Sheet 26)                                            F Al-1 (Sheet 27)                                            F Al-1 (Sheet 28)                                            F Amendment F December 15, 1989        ;
 
li: -    D'                                                      (Shoot 6 of 6) c.
LCESSAREN.
                          ~
      +                          .
EFFECTIVE.PAGE LISTING ~(Cont'd)                    .
APPENDIX A Tables (Cont'd)?                        Amendment Al-1-(Sheet 29)-.                                    F T          .Al-1-(SheetL30).                                        F Al-1: (Sheet'31).                                    F Al-1-(Sheet 32).                                      F-                ,
                      -Al-1-(Sheet 33)                                        F o
                      .Al-11(Sheet 34).                                      F LAl-1-(Sheet ~35).                                      F                a Al-1 (Sheet 3<6)                                      F Al-1 (Sheet;37)                                      F Al-1"(Sheet 38)                                      F Al-1-(Sheet'39)                                      F
                      'Al-1-(Sheet 40)                                        F
                      .Al-1;(Sheet 41)                                        F 1Al-l'(Sheet 42)-                                        F.
L                    JAl-11(Sheet 43)                                        F
                      -Al-1-(Sheet 44)                                        F Al-1'(Sheet 45)                                      F Al-1L(Sheet;46).                                      F LAl-1L(Sheet 47)-                                      F
                      'Al-1 (Sheet-48)                                        F                  i Al-1.:(Sheet 49)-                                    F Al-l'~(Sheet 50)                                      F Al-1 :(Sheet 51)                                      F
      "'                Al-1-(Sheet 52)-                                      F Al-1~(Sheet 52)                                      F sal-l'(Sheet'54)                                      .F
                      .Al-1 (Sheet 55)                                        F A2-1~ (Sheet-1)
F
_A2-1-(Sheet 2)                                        F'
                        .A2-in(Sheet 3)                                        F                  ,
h                        A2-1 (Sheet 4)                                        F
[                      ;A2-11(Sheet 5)                                        F h                        A2-1 (Sheet 6)                                        F                  !
k                        A2-1~(Sheet 7)                                        F l                        A2-1,(Sheet 8)                                        F I
A2-1 (Sheet ~9)                                      F A2-1-(Sheet 10)                                      F L                        A3-1.(Sheet =1)                                      F A3-1~(Sheet 2)                                        F A4-1 (Sheet.1)                                        F A4-l'(Sheet 2)                                        F A4-1 (Sheet 3)                                        F A4-1 (Sheet'4)                                        F A4-1 (Sheet 5)                                        F Amendment F December 15, 1989  i
 
m  3 t- a 1CESSAR' tinbi.*
, 1 F
s APPENDIX A L
CLOSURE OF UNRESOLVED AND GENERIC' SAFETY ISSUES j'
Amendment F      .
December 15, 1989
 
1 l
  ,    CESSAR1!nkm x
APPENDIE A CLOSURE 91 UNRESOLVED AND GENERIC SAFETY 188UE8 OVERVIEW The objective of.this. Appendix is to provide documentation of the technical. resolutions for all Unresolved Safety Issues and
      ' Medium- and High-Priority Generic Safety Issues (USIs and GSIs) which are technically-relevant to the System 80+ Standard Design-as required by 10 CFR Part 52.47 for Design Certification. This Appendix is divided into Sections 1.0 through 4.0 which present listings of the various issues and those applicable to the System 80+ Standard Design.
For this Appendix, 734. USIs and GSIs were evaluated for their applicability to the System 80+ Standard Design based on the review of NRC and industry documentation (e.g., NUREG-0933 "A Prioritization~ of Unresolved and- Generic Safety Issues"          and NUREG-1197, . " Advanced ~ Light Water Reactor Program, Management        -
Review Methodology").        These 734 . issues are listed in Section 1.0. Section_'2.0    identifies the USIs and GSIs technically
        ' applicable to the System 80+ Standard Design and Section . 3.0 contains a cross-reference, among applicable issues, based upon NUREG-0933.      Finally,    Section 4.0 contains the individual technical resolutions to the applicable Unresolved and Generic Safety-Issues.
After  the  System    80+  applicable  issues were  identified,  a methodology.for the documentation of the technical resolution of each issue was developed. Each applicable issue is comprised of four  sections:    ISSUE,  ACCEPTANCE  CRITERIA,  RESOLUTION  and REFERENCES.      The ISSUE statement section ' consists of a brief summary description of the safety issue. This is followed by the ACCEPTANCE CRITERIA section.          These criteria are taken from NUREG-0933 in most cases, and, in the absence of a formal NRC resolution,' developed from accepted industry codes, guidelines, standards and/or good engineering practice. The RESOLUTION section contains the technical resolution of the safety issue which is based upon the System 80+ Standard Design as described
          .within CESSAR-DC or other pertinent documentation as listed in the REFERENCES section. This four part structure for each safety issue is intended to establish how each safety issue is resolved in a clear and concise manner.
Amendment F December 15, 1989
 
i CESSAR Elmace..N TABLE OF.. CONTENTS APPENDIX A Section  Subiect                                      Pace No.
1.0      NRC-LIST OF UNRESOLVED SAFETY ISSUES          A-1 AND GENERIC SAFETY ISSUES (USIs AND GSIs) 2.0      LIST OF UNRESOLVED SAFETY ISSUES AND          A                HIGH/ MEDIUM PRIORITY GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM 80+
STANDARD DESIGN 3.0      APPLICABLE NRC USIs/GSIs CROSS-              A-5 REFERENCED IN NUREG-0933 4.0      TECHNICAL RESOLUTIONS FOR UNRESOLVED          A-7 AND GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN Amendment F i              December 15, 1989
 
          ?,    CESSARinGncmo.                                                                              :
LIST OF TABLES APPENDIX A Table        subiect
              . Al-l'        Listing of Unresolved Safety Issuasiand Generic Safety Issues A2        List of Unresolved Safety Issues and High/ Medium                              ;
Priority Generic Issues Applicable to the System 80+ Standard Design I1 A3-1          NRC USIs/GSIs Cross-referenced in NUREG-0933                                  i and Applicable to the System 80+ Standard                                      !
Design V
1 A4-1          List of Technical Resolutions.for USIs and.GSIs Applicable to the System 80+ Standard Design Included in Section-4.0 l
L I
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    ~
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Amendment F          ,
11                                      December 15, 1989    I l
1
 
CESSAR tm%.v:
1.0        NRC LIST OF UNRESOLVED SAFETY ISSUES AND GENERIC 8AFETY ISSUES (USIs AND GSIs) 1 A  total  of 734    USIs and GSIs      are identified    in  a  "A  l Prioritization of Generic Safety Issues" (NUREG-0933)*, along with a summary of the status of each issue. Of the 734 issues, 38G were considered to be " resolved" on the basis that the EPRI        l Regulatory Stabilization Program, which reviewed all USIs and GSIs, determined them to be "Not Applicable" (see NUREG-1197).
The remaining 348 issues were considered to be applicable to the        ,
design of Advanced Light Water Reactors.        Further review was      l performed by Combustion Engineering, Inc. to determine the subset        i of issues applicable to the System 80+ Standard Design and the following two categories were established:
i Category 1:      Issues not Relevant to the System 80+ Standard Design                                                  ;
An issue was climinated for System 80+ Standard        1 Design if it met one or more of the following
>                        criteria:                                              !
l-                                                                              '
: a. The issue is prioritized in NUREG-0933 as DROPPED or LOW, or the issue has not yet been prioritized.
: b. The  issue is specific    to  another  design (e.g., BWR, H, B&W).
  ,                      c. The NRC identified the issue as resolved with no new requirements and no references to old requirements,
: d. The NRC identified the issue as either an operational, environmental, licensing, or NRC internal issue.
: e. The issue has been superseded by one or more USIs and GSIs.
* The version of NUREG-0933 required by 10 CFR Part 52.47 is        1 that version current six months prior to application.        For the System 80+ Standard Design, the application is dated March 30, 1989 and six months prior to this date, NUREG-0933 through Supplement    8, June 1988, was in effect.      However, for the System 80+ Design Certification Program, NUREG-0933 through Supplement 9, April 1989, was used.
Amendment F A-1                December 15, 1989
 
m [CESSARt h
: f. The issue was classified as a DROP issue in the EPRI Regulatory Stabilization Program (see NUREG-1197).
: g. The issue was classified as NOT APPLICABLE in the  EPRI  Regulatory  Stabilization- Program (see NUREG-1197).
Category 2:    Issues Relevant to the System 80+-standard Design              ;
The above categorie.s and criteria correspond to Categories la                l through ig, and Category 2 in the composite list of issues                      :
presented: in this section. In this list, USIs are " Unresolved Safety Issues" previously identified in the "Unrosolved Safety                  l Issue Summary" NUREG-0606 and referenced in NUREG-0933. GSIs are                j
          " Generic Safety Issues" and, in some cases, they are further classified as a Licensing Issue (GSI/LI) , an NRC Regulatory Impact Issue (GSI/RI), or a Three Mile Island Issue (GSI/TMI).                    .
As resolution of USIs and GSIs progresses through the review                      (
i process, the list-of issues for the System 80+ Standard Design may be revised.
Table Al-1 provides a listing of 734 Unresolved Safety Issues and                  ,
Generic Safety Issues.                                                              i
                                                                                                'I 4
i Amendment F A-2                      December 15, 1989
 
      't CESSAR Ripiccioi,                                                                              [
y                                                                      -
TABLE A1-1 (Sheet 1 of 55)                                                        1 LISTING OF                                                        ,
UNRESOLVED SAFETY _ ISSUES _.AND GENERIQ SAFETY ISSUES
        . ISSUE                                            ISSUE NUMBER              ISSUE TITLE                  TYPE                    CATEGORY
        -1          FAILURES IN                            GSI                    la                    ,.
l AIR-MONITORING, AIR l                    CLEANING, AND VENTILATION l                    SYSTEMS 2        FAILURE OF PROTECTIVE                  GSI                    la DEVICES ON ESSENTIAL EQUIPMENT 3        SETPOINT DRIFT IN
* GSI                    2 INSTRUMENTATION 4        END-OF-LIFE MAINTENANCE                GSI                    id CRITERIA 5        DESIGN CHECK AND AUDIT OF              GSI                    le BALANCE OF PLANT EQUIPMENT 6        SEPARATION OF CONTROL ROD              GSI                    lb TROM ITS DRIVE AND BWR HIGH ROD WORTH EVENTS 7        FAILURES DUE TO                        GSI                    lf FLOW-INDUCED VIBRATIONS 8        INADVERTENT ACTUATION OF              GSI                    le SAFETY INJECTION IN PWRS 9        RE-EVALUATION OF REACTOR              GSI                    le COOLANT PUMP TRIP CRITERIA 10        SURVEILLANCE AND                      GSI                    ig MAINTENANCE OF TIP ISOLATION VALVES AND SQUIB CHARGES 11        TURBINE DISC CRACKING                  GSI                    le Amendment F December 15, 1989
 
1 4
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y LCESSAR Mincues TABLE Al-1 (Cont'd)                                                                      ,
(Shect 2 of 55)
LISTING OF UNRESOLVED SA*ETY ISSUES AND                                                                'i GENERIC SAFETY ISSUES ISSUE'                                                            ISSUE-NUMBER                    ISSUE' TITLE                            TYPE                            CATEGOBX 12      BWR JET PUMP INTEGRITY                                    GSI                              lb-13    SMALL BREAK LOCA TROM-                                      GSI                              if EXTENDED OVERHEATING OF-PRESSURIZER HEATERS 14'-    PWR PIPE CRACKS                                            GSI                              2 15    RADIATION EFFECTS ON                                        GSI                              2 REACTOR VESSEL SUPPORTS 16      BWR, MAIN STEAM ISOLATION.                                GSI                              if VALVE LEAKAGE CONTROL ll                  SYSTEMS-                                                                                                .!
L l'            17      LOSS OF OFFSITE POWER                                      GSI                              lf            '
SUBSEQUENTIAL SMALL LOCA L            18    STEAM LINE BREAK WITH                                      GSI                              le CONSEQUENTIAL SMALL LOCA 19    SAFETY IMPLICATIONS OF                                      GSI                              le NON-SAFETY INSTRUMENT AND CONTROL POWER SUPPLY BUS 20    EFFECTS OF ELECTROMAGNETIC                                  GSI                              1c PULSE ON NUCLEAR PLANT SYSTEMS 21    VIBRATION QUALIFICATION OF                                  GSI                              lf EQUIPMENT 22    INADVERTENT BORON DILUTION                                  GSI                              2 EVENTS Amendment F December 15, 1989
 
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          +
                            . CESSAR Mnco.
l 1
TABLE A1-1 (Cont'd)
(8heet 3 of 55)
  ,                                                                                                                                                                      l '
LISTING OF UNRESOLVED SAFETY ISSUE 8 AND GENERIC SAFETY ISSUES ISSUE:                                                                                            ISSUE NUMBER                            ISSUE TITLE                                      -
TYPE      CATEGORY                      !
23                        REACTOR COOLANT PUMP SEAL                                                GSI      2                              ,
FAILURES 24                        AUTOMATIC EMERGENCY CORE                                                GSI      la                            l COOLING SYSTEM SWITCH TO                                                                                          >
RECIRCULATION                                                                                            ,
t j                              25                        AUTOMATIC AIR HEADER DUMP                                                GSI      lb ON BWR SCRAM SYSTEM 26                        DIESEL GENERATOR LOADING                                                GSI      le PROBLEMS RELATED TO SIS                                                                                -
RESET ON LOSS OF OFFSITE POWER 27                        MANUAL VERSUS' AUTOMATED                                                GSI      le                  [
ACTIONS 25                        PRESSURIZED THERMAL SHOCK                                                GSI      le 1                              29                        BOLTING DEGRADATION OR                                                  GSI      2 FAILURES-IN NUCLEAR PLANTS 30                        POTENTIAL GENERATOR                                                      GSI      if                              )
L              ,                                        MISSILES -- GENERATOR                                                                                              j H                                                        ROTOR RETAINING RING                                                                                                ,
I 31                        NATURAL CIRCULATION                                                      GSI      le COOLDOWN L                              32                        FLOW BLOCKAGE IN                                                        GSI      le ESSENTIAL EQUIPMENT CAUSED BY CORBICULA l
l Amendment F December 15, 1989
      --#-d -  ---a    _ *    *__-.A  a__ _____ _- _      __        m  - r- wr.--.> , sw..w, -.a m.yee--waw---  -._%gw-rwy-        ,,    -.,wy e.c ,+g ,,    g- ,,
 
  -                                                                                                    -- ~
n    CESSAR ti!th..                                                                                              :
1 TABLE 11-1 (Cont #d)
(Sheet 4 of 55) s LISTING OF UMRESOLVED 4AFETY ISSUES AND GEMERIC SAFETY ISSUES ISSM                                                                ISSUE 303333                      ISSUE TITLE                            TYPE      CATEGORY 33            CORRECTING ATMOSPHERIC                                GSI      le                            ;
DUMP VALVE OPENING UPON LOSS OF INTEGRATED CONTROL SYSTEM POWER-3$            REACTOR COOLANT SYSTEM                                GSI      1f                            -
LEAK 35            DEGRADATION OF INTERNAL                              GSI      la APPURTENANCES IN LWR'S                                                                -
36            LOSS OF SERVICE WATER                                GSI      2 37            STEAM GENERATOR OVERFILL                              GSI      le
                    & COMBINED PRIMARY &
SECONDARY BLOWDOWN                                                                            ,
38            POTENTIAL RECIRCULATION                              GSI      lb FAILURE AS A CONSEQUENCE OF CONTAINMENT PAINT OR DEBRIS 39            POTENTIAL FOR UNACCEPT                                GSI      lb INTERACTION BETWEEN THE CRD SYSTEM & NON-ESSENTIAL CONTROL AIR SYSTEM (BWR) 40            SAFETY CONCERNS                                      GSI      lb ASSOCIATED WITH BREAKS IN THE BWR SCRAM SYSTEM 41            SCRAM DISCHARGE VOLUME                                GSI      lb SYSTEMS 42            COMBINATION PRIMARY &                                GSI      le SECONDARY SYSTEM LOCA Amendment F December 15, 1989
 
o o.
CESSARlR hi.
i TABLE 11-1 (Cont *4)
(sheet 5 of 55)
LISTING OF mrgg30LvaD s&FETY ISSUES AMD gnummIc sAFaTY ISSUES Issus                                    Issus MUMBER            ISSUE TITLE          , TYPE      CATEGORY 43      CONTAMINATION OF-                GSI      lf INSTRUMENT AIR LINES 44      FAILURE OF SALT WATER            GSI      la COOLING 45      INOPERABILITY OF                  GSI      2 INSTRUMENTS DUE TO EXTREME COLD WEATHER 46    ICSS OF 125 VDC BUS              GSI      le 47    THE LOSS OF OFFSITE POWER        GSI      le 48      LCO FOR CLASS 1E VITAL            GSI      2 INSTRUMENT BUSES IN OPERATING REACTORS 49      INTERI4CKS AND LCOs FOR          GSI      2 REDUNDANT CLASS 1E TIE BREAKER 50      REACTOR VESSEL LEVEL IN          GSI      lb BWRS 51      PROPOSED REQUIREMENTS FOR        GSI      2 IMPROVING RELIABILITY OF OPEN CYCLE SERVICE WATER SYSTEMS 52      FLOW BLOCKAGE BY BLUE            GSI        le MUSSELS 53      CONSEQUENCES OF A              _GSI        lf POSTULATED BICCKAGE INCIDENT IN A BWR l'
Amendment F December 15, 1989
 
n                                                                                                      1 CESSAR tinhie                                                                                      l l
l TABLE 11-1 (Cont'd)
(Sheet 6 of 55)                                                                ,
LISTING _QE UMmESOLVED BAFETY ISSUES AND                                                          :
GIFERIC SAFETY ISSUES ISSUE                                                    ISSUE MUMBER          ISSUE TITLE                            TYPE.                          CATEGORY 54    VALVE OPERATOR RELATED                            GSI                          le EVENTS OCCURRING DURING l            '78, '79, 80 l                                                                                                        .
55    FAILURE OF CLASS 1E                              GSI                            lf        '
SAFETY RELATED SWITCHGEAR CIRCUIT BREAKER TO CLOSE ON DEMAND 56    ABNORMAL TRANSIENT                              GSI                            le OPERATING GUIDELINES AS APPLIED TO STEAM GENERATOR OVERFILL EVENT 57    EFFECTS OF FIRE                                  GSI                          2 l            PROTECTION SYSTEM                                                                          ,
ACTUATION ON SAFETY                                                                        .
RELATED EQUIPMENT                                                                          -
L 58    INADVERTENT CONTAINMENT                          GSI                          lf FLOODING l
l      59    TECHNICAL SPECIFICATION                          GSI/RI                        id        ,
REQUIREMENTS FOR PLANT SKUTDOWN 60    LAMELLAR TEARING OF RCS                          GSI                            le l            STRUCTURAL PARTS 61    SRV DISCHARGE LINE BREAK                        GSI                            lb l            INSIDE TO WETWELL l            AIRSPACE OF MARK I & III CONTAINMENT I
Amendment F i                                                                                      December 15, 1989 l
 
0-CESSAR IMWcui..
I f
TABLE 11-1 (Cont'd)                                                ;
(sheet '' of SS)                                            .
LISTING OF                                                l UNRESOLYSD SAFETY Issues AND                                                            l eEMERIc shrETY Issues                                                  ;
issue                                    i Issue Mgg333                      ISSUE TITLE                    TYPE          C&TEGORY 62      REACTOR SYSTEMS BOLTING                            GSI          le APPLICATIONS l
63      USE OF EQUIPMENT NOT                              GSI          lb                        l CLASSIFIED AS ESSENTIAL TO SAFETY IN BWR                                                                        ._t TRANSIENT ANALYSIS                                                                          i 64      IDENTIFICATION OF                                  GSI          2 PROTECTION SYSTEM                                                                          -
INSTRUMENT SENSING LINES                                                                    ,
65      PROBABILITY OF CORE MELT                          GSI          le DUE TO COMPONENT COOLING -                                                                  ,
WATER SYSTEN FAILURES 66      STEAM GENERATOR                                    GSI          2 REQUIREMENTS 67.2.1  STEAM GENERATOR STAFF                              GSI/RI        ld l              ACTIONS--INTEGRITY OF l              STEAM GENERATOR TUBE l              SLEEVES 67.3.1  STEAM GENERATOR STAFF                              GSI          le ACTIONS--STEAM GENERATOR OVERFILL 67.3.2. STEAM GENERATOR STAFF                              GSI          le ACTION--PRESSURIZED                                                                        ,
l              THERMAL SHOCK 67.3.3  STEAM GENERATOR STAFF                              GSI/LI        le ACTIONS -- IMPROVED ACCIDENT HONITORING Amendment F December 15, 1989
 
M CESSARHanew.                                                                                  !
I TABLE 11-1 (Cont'd) l (Sheet 8 of 55)
LISTING OF                                            ;
UNRESOLVED SAFETY ISSUES AND GENERIC ShFSTY ISSUES                                              l ISSUE                                                                ISSUE NUMBER                    -ISSUE TITLE                              TYPE          CATEGORY
* 67.3.4      STEAM GENERATOR STAFF                                    GSI          le ACTIONS--REACTOR VESSEL INVENTORY MEASUREMENT 67.4.1      STEAM GENERATOR STAFF                                  GSI          le ACTIONS--REACTOR COOLANT PUMP TRIP                                                                        ,
67.4.2      FTEAM GENERATOR STAFF                                  GSI          la ACTIONS--CONTROL ROOM DE' SIGN REVIEW 67.4.3      STEAM GENERATOR STAFF                                  GSI          le        i ACTIONS--EMERGENCY OPERATING ?ROCEDURES                                                            ,
67.5.1      REASSESSMENT OF STGR                                    GSI/LI        id DESIGN BASIS j .,                  (RADIOLOGICAL l
CONSEQUENCES) 67.5.2      REEVALUATION OF SGTR                                    GSI/LI        id DESIGN BASIS l      67.5.3      STEAM GENERATOR STAFF                                  G3I          if ACTIONS--SECONDARY SYSTEM ISOLATION 67.6.0      STEAM GENERATOR STAFF                                  GSI          le l
ACTION--ORGANIZATIONA's
 
===RESPONSE===
67.7.0      STEAM GENERATOR STAFF                                  GSI          le ACTIONS--EDDY CURRENT TESTS 67.8.0      STEAM GENERATOR STAFF                                  GSI/RI        id ACTION--DENTING CRITERIA l
L                                                                              Amendment F December 15, 1989 l
 
i CESSAR !!mam..
I 1
i TAELE 11-1 (Cont'd)                              l i
                    <S.... . ., 883 LISTlMG OF                                    ,
UMmESOLVED SAFETY ISSUES AND                            l aEMEmIe ShrETY ISSUES                              j ISSUE                                        ISSUE                  l NUMEER            ISSDE TITLE                TYPE      CATEGORY      )
1 67.9.0  STEAM GENERATOR STAFF              GSI      le            I ACTION--REACTOR COOLANT                                      l SYSTEM PRESSURE CONTROL 67.10.0  STEAM GENERATOR STAFF              GSI/LI    id            )
ACTION--SUPPLEMENTAL TUBE INSPECTIONS 68      POSTULATED LOSS OF AFWS            GSI      le            !
RESULTING FROM TURBINE                                      j DRIVEN AFW PUMP STEAM SUPPLY LINE BREAK 69      MAKE-UP NOZZLE CRACKING            GSI      lb IN B&W PLANTS 70      PORV AND LOCK VALVE                GSI      lb            I l
RELIABILITY 71      FAILURE OF RESIN                    GSI      la DEMINERALIZER SYSTEMS AND THEIR EFFECT ON PLANT SAFETY 72      CONTROL ROD DRIVE GUIDE            GSI      la            i TUBE SUPPORT PIN FAILURES 73      DETACHED THERMAL SLEEVES            GSI      lf 74      REACTOR COOLANT ACTIVITY            GSI      la LIMIT FOR OPERATING REACTORS 75      GENERIC IMPLICATIONS OF            GSI      2 ATWS EVENTS AT SALEM --
OPERATIONAL QA PROGRAMS Amendment F December 15, 1989
 
1
    .-  CESSAR !!ntricui .                                                                                                                                  l i,
i TAELE A1-1 (Cont'd) i t
l                                                    (Sheet 10 of 55)                                                                                        J l-                                                                                                                                                            ;
I                                                                LISTING OF
: i.                                            UNRESOLVED SAFETY ISSUES.AND GENERIC SAFETY ISSUES ISSUE                                                                                                ISSUE l          NUMEER                                  ISSUE TITLE                                                  TYPE._                    CATEGORY 76                            INSTRUMENTATION AND                                                    GSI                        la                l CONTROL POWER INTERACTIONS 77                            FLOODING OF SAFETY                                                    GSI                        le EQUIPMENT COMPARTMENTS BY                                                                                          .)
BACKFLOW I          78                            MONITORING OF FATIGUE                                                  GSI                        lf l
TRANSIENT LIMITS FOR                                                                            '
REACTOR COOLANT SYSTEM                                                                                    .        l l
79                            UNANALYZED REACTOR VESSEL                                              GSI                        2                l THERMAL STRESS- DURING NATURAL CONVECTION COOLDOWN 80                            PIPE BREAK EFFECTS ON CRD                                              GSI                        lb HYDRAULICS IN BWR MK I &                                                                                            ]
II CONTAINMENTS                                                                                                    l 81                            POTENTIAL PROBLEMS WITH                                                GSI                        lf I4CKED DOORS & BARRIERS                                                                                            l IN NUCLEAR PLANTS                                                                                                  l 82                            BEYOND DESIGN BASES                                                    GSI                        2 ACCIDENTS IN SPENT FUEL                                                                                            ;
POOLS 83                            CONTROL ROOM HABITABILITY                                              GSI                        2 1
84                            C-E PORVS                                                              GSI                        la
;          85                            RELIABILITY OF VACUUM                                                  GSI                        lb BREAKER INSIDE BWR I                                        CONTAINMENTS Amendment F December 15, 1989
 
C E S S A R iintneu ..r                                                                                              1 TABLE 11-1 (Cont'd)                                                              -
(sheet 11 of 55)                                                                    I 1
LISTING OF                                                                    )
UmmasOLvan sarzTT Issuns hND                                                            l annamIC saraTY Issuns                                                                j
                                                                                                                            ]
issue                                                                            issue                                  i MUMBER                            ISSUE TITLE                                  TYPE      CATEGORY                  -
1 86                      LONG RANGE PLAN FOR                                      GSI      lb I
DEALING WITH SSC IN BWR                                                                        j PIPING i    B7                      FAILURE OF HPCI STEAM-                                  GSI      lb LINE WITHOUT ISOLATION (IN BWR'S) l L    88                      EARTHQUAKE AND EMERGENCY                                GSI      lg i                            PLANNING                                                                                      ,
89                      STIFF PIPE CLAMPS                                        GSI      lf 90                      TECH. SPECS. FOR                                        GSI      1g ANTICIPATORY TRIPS 91                      MAIN CRANKSHAFT FAILURE                                  GSI      1c                          '
IN TRANSAMERICA DELAVAL EMERGENCY DIESEL GENERATORS 92                      FUEL CRUMBLING DURING                                    GSI      la LOCA 93                      STEAM BINDING OF                                        GSI      2 AUXILIARY FEEDWATER PUMPS 94                      ADDITIONAL LTOP FOR LIGHT                                GSI      2 WATER REACTORS                                                                                ,
t 95                      LOSS OF EFFECTIVE VOLUME                                GSI      lb FOR CONTAINMENT RECIRCULATION 96                      RHR SUCTION VALVE TESTING                                GSI      lb 97                      PWR REACTOR CAVITY                                      GSI      le UNCONTROLLED EXPOSURE Amendment F December 15, 1989
 
Q)
CESSAR WWienion                                                      :
TABLE 11-1 (Cont'd)                                ,
(Sheet 12 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND I
GENERIC SAFETY ISSUES ISSUE                                      ISSUE MUMBER          ISSUE TITLE                TYPE        CATEGORY 98      CRD ACCUMULATOR CHECK              GSI          lb VALVE LEAKAGE (IN BWR'S)                                      .
99      RCS/RHR SUCTION LINE                GSI          2 INTERLOCKS ON PWRS 100    OTSG LEVEL (B&W)                    GSI          lb 101    BWR WATER LEVEL                    GSI          lb REDUNDANCY 102    HUMAN ERROR IN EVENTS              GSI          le INVOLVING WRONG UNIT OR WRONG TRAIN 103    DESIGN FOR PROBABLE                GSI          2 MAXIMUM PRECIPITATION 104    REDUCTION OF BORON                  GSI/RI      id DILUTION REQUIREMENT 105    INTERFACING SYSTEMS LOCA            GSI          2 AT LWRS 106    PIPING AND USE OF HIGHLY            GSI          2 COMBUSTIBLE GASES IN VITAL AREAS -- FIRE PROTECTION 107    GENERIC IMPLICATIONS OF            GSI          la MAIN TRANSFORMER FAILURES 108    BWR SUPPRESSION POOL                GSI/RI      lb LIMITS 109    REACTOR VESSEL CLOSURE              GSI          lb            ;
FAILURE l-Amendment F December 15, 1989
 
CESSARBR Ln..
TABLE 11-1 (Cont'd)
                  -(sheet 13 of 55)
LISTING OF UMRESOLVED BAFETY ISSUER AND GENERIC SAFETY ISBUES ISSUE                                    ISSUE MUMBER          ISSUE TITLE            TYPE.      CATEGORY 110    EQUIPMENT PROTECTION            GSI        lb DEVICES ON ENGINEERED SAFETY FEATURES 111    SCC OF PRESSURE BOUNDARY        GSI/LI    ld FERRETIC STEELS IN SELECTED ENVIRONMENTS 112    WESTINGHOUSE RPS                  GSI/RI    lb SURVEILLANCE FREQUENCIES
        & OUT-OF-SERVICE TIMES 113    DYNAMIC ' QUALIFICATION          GSI      le TESTING OF LARGE ORE HYDRAULIC SNUBBERS 114    SEISMIC' INDUCED RELAY            GSI        le CHATTER 115    ENHANCEMENT OF THE                GSI        lb RELIABILITY OF THE WESTINGHOUSE SSPS (ATWS) 116    ACCIDENT MANAGEMENT              GSI        la 117    ALLOWABLE OUTAGE TIMES FOR      GSI        la DIVERSE SIMULTANEOUS EQUIPMENT OUTAGES 118    TENDON ANCHORAGE FAILURE        GSI        la 119.1  PIPE RUPTURE REQUIREMENTS        GSI/RI    2 119.2  PIPE DAMPING VALUES              GSI/RI    2 119.3  DECOUPLING OBE FROM SSE          GSI/RI    2 119.4  BWR PIPING MATERIALS              GSI/RI    lb l
Amendment F December 15, 1989
 
I I
k !$lhlh klI $bN f6CATl0N                                                          i l
l l
TkBLE 11-1 (Cont'd)
(Sheet 14 of 55)                                      l LISTING OF UMRESOLVED SAFETY ISSUES AND                                  ,
GENERIC M&FETY IPSUES                                    l
                                                                                    )
issue                                                        ISSUE MUMBER                    ISSUE TITLE                        TYPE    CATEGORY 119.5            LEAK DETECTION                              GSI/RI  2 REQUIREMENTS 120              ON-LINE TESTABILITY OF                      GSI      la PROTECTION SYSTEMS 121              HYDROGEN CONTROL FOR                        GI.      2 LARGE, DRY PWR CONTAINMENTS                                                    4 122.la          COMMON MODE FAILURE OF                      GSI      le ISOLATION VALVES IN CLOSED POSITIONS 122.lb          RECOVERY OF AUXILIARY                        GSI      le FEEDWATER I
122.lc          INTERRUPTION OF AUXILIARY                    GSI      le FEEDWATER FLOW 122.2            INITIATING FEED AND BLEED                    GSI      2 122.3            PHYSICAL SECURITY SYSTEM                    GSI      la        ,
CONSTRAINTS 123              DEFICIENCY IN THE                            GSI      la        ,
REGULATIONS GOVERNING DBA l                  AND SINGLE FAILURE
[                  CRITERION - DAVIS BESSE l
124              AUXILIARY FEEDWATER                          GSI      2 SYSTEM RELIABILITY 125.I.1          AVAILABILITY OF THE                          GSI      la l                  SHIFT TECHNICAL ADVISOR l
125.I.2.a        PORV RELIABILITY                            GSI      le Amendment F December 15, 1989
 
t i
CESSAR tlELn..                                                                        ;
f f
TABLE 11-1 (Cont'd)
                          <S .. 1. .,..>                                            !
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC BAFETY ISSUES ISSUE                                          ISSUE MUMBER                ISSUE TITLE        _
TYPE      CATEGORY 125.I.2.b    POR\' EELIABILITY                GSI      le                          ;
125.I.2.c    AUTO LOCK VALVE CLOSURE          GSI      lf 125.I.2.d    EQUIPMENT QUALIFICATION          GSI      le FOR FEED & BLEED                                                    --.
ENVIRONMENT 125.I.3      SPDS AVAILABILITY                GSI      2 125.I.4      LONG TERM ACTIONS -              GSI      la                          j DAVIS BESSE EVENT--PLANT                      ,
SPECIFIC SIMULATOR                                                      ;
125.I.5      SAFETY SYSTEM TESTED IN          GSI      la ALL CONDITIONS REQUIRED BY DESIGN BASIS ANALYSIS 125.I.6      VALVE TORQUE LIMIT AND            GSI      la BYPASS SWITCH SETTINGS                                                  ',
1 125.I.7.a    RECOVER FAILED EQUIPMENT          GSI      la 125.I.7.b    REALISTIC HANDS ON                GSI      la TRAINING 125.I.8      PROCED'JRES AND STAFFING          GSI      la FOR REPORTING TO NRC EMERGENCY RESPONSE CENTER 125.II.I.a  TWO-TRAIN AFW RELIABILITY        GSI      la 12.II.1.b    REVIEW EXISTING AFWS FOR          GSI      le SINGLE FAILURE 125.II.1.c  NUREG-0737 RELIABILITY            GSI      la IMPROVEMENTS l
l l                                                  Amendment F l                                                  December 15, 198P l
 
1 CESSAR ilnincua                                                                                            l J
TABLE Al-1 (Contid)
(Sheet is of 55)                                                        .j LISTING CF UNRESOLVED SAFETY ISSUES AND 9ENERIC SAFETY ISSUES ISSUE                                                              ISSUE                                  ;
NUMBER                      188DB TITLE                            Z1 R        CATEGORY 125.II.1.d AFW STEAM & FEED RUPTURE                                GSI        la CONTROL SYSTEM /ICS INTERACTION IN B&W PLANTS 125.II.2  ADEQUACY OF EXISTING                                    GSI        la MAINTENANCE REQUIREMENTS FOR SAFETY RELATED SYSTEMS 125.II.3 , REVIEW STEAM / FEED LINE                                GSI        la                        r BREAK SYSTEMS FOR SINGLE                                      ,
FAILURE 125.II.4  OTSG DRYOUT & REFLOOD                                    GSI        lb                      .I EFFECT (B&W)                                                                                    ,
125.II.5  THERMAL-HYDRAULIC                                        GSI        la EFFECT-LOSS AND RESTORATION OF FDW ON PRIMARY SYSTEM COMPONENTS                                                                      ,.
125.II.6  REEXAMINE PRA ESTIMATE OF                                GSI        la CORE DAMAGE RISK FROM LOSS OF FEEDWATER 125.II.7  REEVALUATE PROVISION TO                                  GSI        2 AUTOMATICALLY ISOLATE                                                                          ,
FEEDWATER FROM STEAM GENERATOR DURING LINE BREAK 125.II.8  REASSESS CRITERIA FOR                                    GSI        la FEED & BLEED INITIATION 125.II.9  ENHANCED FEED & BLEED                                    GSI        le CAPABILITY Amendment F December 15, 1989
 
t CESSAR BI!#iemos l
TABLE 11-1 (Cont'd)
(sheet 17 of 55)                                                      i LIsTIme or minnsOLVan sAraTY Issung nun ammanIC sarzTY Issess issue                                                              issue                                  +
NUMBER                          ISBUE TITLE                        TYPE            CATEGORY 125.II.10          HIERARCHY OF IMPROMPTU                          GSI              la
(.                    OPERATOR ACTIONS I
;  125.II.11          RECOVERY OF MAIN                                GSI              la                    >
FEEDWATER AS AN ALTERNATIVE TO AFW i  125.II.12          LONG-TERM GENERIC ACTIONS                      GSI              id AS A RESULT OF THE DAVIS-EESSE EVENT OF 6/9/5--ADEQUACY OF TRAINING REGARDING PORV OPERATION 125.II.13          OPERATOR JOB AIDS                              GSI              la                    .
125.II.14          REMOTE OPERATION OF                              GSI              la l
EQUIPMENT WHICH MUST NOW BE OPERATED IDCALLY                                                                    .
126                RELIABILITY OF PWR MAIN                          GSI/LI          id STEAM SAFETY VALVES 127                TESTING AND MAINTENANCE                          GSI              la OF MANUAL VALVES IN SAFETY RELATED SYSTEMS l'
E 128                ELECTRICAL POWER                  <
GSI              2                    -
RELIABILITY l  129                VALVE INTERLOCKS TO                              GSI              le l                    PREVENT VESSEL DRAINAGE l                    DURING SHUTDOWN COOLING l'
130                ESSENTIAL SERVICE WATER                          GSI              2 PUMP FAILURE AT MULTIPLANT SITES Amendment F December 15, 1989
 
C E S S A R 88r' % m .                                                                                                    .
T&BLE 11-1 (Cont'd)                                                            :
(Sheet is of 55)
LISTING OF                                                                  ;
UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES                                                            ,
ISMUE                                                            ISSUE                                                  ;
NUMBEg                                  ISSUE TITLE              TYPE            CATEGORY 131          POTENTIAL SEISMIC                                  GSI              lb INTERACTION INVOLVING THE MOVABLE INCORE FLUX MAP SYSTEM AT WESTINGHOUSE PIANTS 132          RHR PUMPS INSIDE                                  GSI              la CONTAINMENT                                                                                              ,
133          UPDATE POLICY STATEMENT                            GSI/LI          id ON NUCLEAR PLANT STAFF WORKING HOUR 134          DEGREE AND EXPERIENCE                              GSI              id REQUIRED FOR SENIOR OPERATORS 135          INTEGRATED STEAM                                  GSI              2 GENERATOR ISSUES 136          STORAGE AND USE OF LARGE                          GSI/LI          ld QUANTITIES OF CRYOGENIC COMBUSTIBLES 137          REFUELING CAVITY SEAL                              GSI              la FAILURES 138          DEINERTING UPON DISCOVERY                          GSI              lb OF RCS LEAKAGE 139          THINNING OF CARBON STEEL                          GSI              lb PIPING IN LWRS 140          FISSION PRODUCT REMOVAL                            GSI              lb BY CONTAINMENT SPRAYS OR POOL Amendment F December 15, 1989
 
CESSAR !RrT, car.
TABLE 11-1 (Cont'd)
(Sheet 19 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE                                      ISSUE NUMBER              ISBUE TITLE            TYPE      CATEGORY 141      LBLOCA WITH CONSEQUENTIAL        GSI      la SGTR 142      LEAKAGE THROUGH                  GSI      la ELECTRICAL ISOLATORS
,  143      AVAILABILITY OF CHILLED          GSI      la WATER SYSTEMS 144      SCRAM WITHOUT A                  GSI      la        ;
TURBINE / GENERATOR TRIP 145      IMPROVE SURVEILLANCE AND          GSI      lb STARTUP TESTING PROGRAMS
,  A-1      WATER HAMMER                      USI      2 l
A-2      ASYMMETRIC BLOWDOWN LOADS        USI      2          -.
ON RCS A-3      WESTINGHOUSE STEAM                USI      lb GENERATOR TUBE INTEGRITY                                  l A-4      C-E STEAM GENERATOR TUBE          USI      2 INTEGRITY
: l. A-5      B&W STEAM GENERATOR TUBE          USI      lb INTEGRITY A-6      MARK I SHORT-TERM PROGRAM        USI      lb A-7      MARK I LONG-TERM PROGRAM          USI      lb A-8      MARK II CONTAINMENT POOL          USI      lb DYNAMIC LOADS--LONG TERM PROGRAM Amendment F December 15, 1989 i
J
 
o CESSAR WWcoS TABLE 11-1 (Coat *d)
(Sheet to of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE                                                          ISSUE NUMBER                  ISSUE TITLE                            TYPE          CATEGORY A-9          ANTICIPATED TRANSIENTS WITHOUT                    USI          2        >
SCRAM (ATWS)
A-10          BWR FEEDWATER NOZZLE                              USI          lb CRACKING A-11          REACTOR VESSEL MATERIAL                          USI          lb TOUGHNESS A-12          FRACTURE TOUGHNESS OF                            USI          2 STEAM GENERATOR & RCP SUPPORTS A-13          SNUBBER OPERABILITY                              GSI          2 ASSURANCE                                                                i A-14          FLAW DETECTION                                    GSI          if A-15          PRIMARY COOLANT SYSTEM                            GSI          2 DECONTAMINATION AND STEAM GENERATOR CHEMICAL CLEANING 1
A-16          STEAM EFFECTS ON BWR CORE                        GSI          lb SPRAY DISTRIBUTION A-17          SYSTEMS INTERACTION                              USI          2 i  A-18          PIPE RUPTURE DESIGN                              GSI          if CRITERIA A-19          DIGITAL COMPUTER                                  GSI          id PROTECTION SYSTEM                                                        l A-20          IMPACTS OF THE COAL FUEL                          GSI          Id CYCLE l
Amendment F December 15, 1989
 
l CESSAR !B&m.S                                                                      i I
l TABLE 11-1 (Cont'd)                                            ]
L                            (Sheet 31 of 55)
LISTING OF                                                )
UMRESOLVED SAFETY ISSUES AMD                                        <
GENERIC BAFETY, 233033 ISSUE                                              ISSUE MUMEEE            ISSUE. TITLE                    TYPE      CATEGORY A-21      MAIN STEAMLINE BREAF                      GSI      la INSIDE CONTAINMENT--
EVALUATION OF ENVILONMENTAL CONDITIONS FOR EQU1PMENT QUALIFICATION A-22      PWR MAIN STEAMLINE                        GSI      lf          ;
BREAK--CORE, REACTOR VESSEL AND CONTAINMENT BUILDING RESPONSE                                                .
A-23      CONTAINMENT LEAK TESTING                  GSI/RI    ld A-24      QUALIFICATION OF CLASS 1E                USI      2 SAFETY RELATED EQUIPMENT A-25      NON-SAFETY LOADS ON CLASS                GSI      2                  .
1 1E POWER SOURCES A-26      REACTOR VESSEL PRESSURE                  USI      2 TRANSIENT PROTECTION A-27      REIDAD APPLICATIONS                      GSI/LI    id A-28      INCREESE IN SPENT FUEL                    GSI      1c                '
CAPACITY A-29      PLANT DESIGN FOR                          GSI      2 REDUCTION OF VULNERABILITY TO SABOTAGE                                            .
A-30      ADEQUACY OF SAFETY                        GSI      2 RELATED DC POWER SUPPLIES                                          ,
A-31      RHR SHUTDOWN REQUIREMENTS                USI      2 A-32      MISSILE EFFECTS                          GSI      le Amendment F December 15, 1989 i
 
CESSAR !!nkr..S                                                                                                  .
i TABLE 11-1 (Cont'd)                                                                ;
(Sheet 23 of 55)
LISTING OF                                                                      !
UNRESOLVED SAFETY ISSUES AND                                                                ,
GENERIC SAFETY ISSUES
  -ISSUE                                                                                    ISSUE NUMBER                      ISSUE TITLE                                                  TYPE        CATEGORY A-33            NEPA REVIEW OF ACCIDENT                                                GSI        ld RISKS                                                                                          i A-34            INSTRUMENTS FOR                                                        GSI        le MONITORING RADIATION AND PROCESS VARIABLES DURING ACCIDENTS A-35            ADEQUACY OF OFFSITE POWER                                              GSI        2            ;
SYSTEMS                                                                                          ,
                                                                                                                  .l A-36            CONTROL OF HEAVY LOADS                                                  USI        2          ,
NEAR SPENT FUEL                                                                                  )
A-37            TURBINE MISSILES                                                        GSI        if A-38            TORNADO MISSILES                                                        GSI        la A-39            DETERMINATION OF SAFETY                                                USI        lb RELIEF VALVE POOL DYNAMIC LOADS (IN BWR'S) 1 A-40            SEISMIC DESIGN--SHORT                                                  USI        2            l TERM PROGRAM                                                                                    j A-41            LONG TERM SEISMIC PROGRAM                                              GSI        le          l I
A-42            PIPE CRACKS IN BOILING                                                  USI        lb          l WATER REACTORS                                                                                    l A-43            CONTAINMENT EMERGENCY                                                  USI        2 SUMP PERFORMANCE                                                                                l A-44            STATION BLACKOUT                                                        USI        2 i
A-45            SHUTDOWN DECAY HEAT                                                    USI        2 REMOVAL REQUIREMENTS Amendment F December 15, 1989
 
i l
C E S S A R II M ,c m .,
i l
TABLE A1-1 (Cont'd)                                l (Sheet 23 of 55)
LISTING OF                                  l UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES                                !
IsscE                                              Issue                    l NUMBER                ISSUE TITLE                TYPE    CATEGORY A-46          SEISMIC QUALIFICATION OF            USI      ld              i EQUIPMENT IN OPERATING                                      l PIANTS i
A-47          SAFETY IMPLICATIONS OF              USI      2              j CONTROL SYSTEMS A-48          HYDROGEN CONTROL,                  USI      2 MEASURES & EFFECT OF HYDROGEN BURNS                                          -
A-49          PRESSURIZED THERMAL SHOCK          USI      2 B-1            ENVIRONMENTAL TECHNICAL            GSI      ld SPECIFICATIONS B-2            FORECASTING ELECTRICITY            GSI      ig        ~
i DEMANDS B-3            EVENT CATEGORIZATION                GSI      id B-4            ECCS RELIABILITY                    GSI      le B-5            DUCTILITY OF TWO-WAY                GSI      2 SIABS & SHELLS -- STEEL                                    !
CONTAINMENTS B-6            LOAD, LOAD COMBINATIONS,            GSI      le STRESS LIMITS B-7          SECONDARY ACCIDENT                  GSI/LI  ld CONSEQUENCE MODELING B-8          LOCKING OUT OF ECCS                GSI      lf      ,
POWER-OPERATED VALVES i
Amendment F December 15, 1989    ;
 
J CESSAR Elinnemos                                                    l l
I TmBLE mi-1 (Cont'd)                            {
(sheet 24 of 55)
LIsTIMa OF                                ,
unansOLvan sarzTr Issons nun                        !
azuzmIc saraTY Issess                          .
i ISSUE                                          ISSUE NUMBER              Issue TITLE                TYPE      CATEGORY  >
B-9        ELECTRICAL CABLE                    GSI        1c PENETRATIONS OF                                          ,
CONTAINMENT B-10      BEHAVIOR OF BWR MARK III            GSI        lb CONTAINMENTS B-11      SUBCOMPARTMENT STANDARD            GSI/LI    id        :
PROBLEMS B-12    . CONTAINMENT COOLING              -
GSI        1c        ,
REQUIREMENTS (NON-lhCA)
B-13      MARVIKEN TEST DATA                  GSI/LI    id EVALUATIONS B-14      STUDY OF HYDROGEN MIXING            GSI        le
* CAPABILITY IN CONTAINMENT POST-LOCA                                                ,
B-15      CONTEMPT COMPUTER CODE              GSI/LI    id MAINTENANCE                                              ,
B-16      PROTECTION AGAINST                  GSI        le POSTULATED PIPING FAILURES IN FLUID SYSTEMS OUTSIDE CONTAINMENT                                      ,
l    B-17      CRITERIA FOR SAFETY                GSI        le l              RELATED ACTIONS B-18      VORTEX SUPPRESSION                  GSI        le
!-              REQUIREMENTS FOR l              CONTAINMENT SUMPS B-19      THERMAL-HYDRAULIC                  GSI        ic STABILITY Amendment F December 15, 1989
 
l
, CESSAR RBWicar.
TABLE 11-1 (Cont'd)
(Sheet 35 of 55)
LISTIMS OP UMERSOLVED R&FETY 188UE8 &MD GENERIC R&FETY ISSUES ISSUE                                            ISSUE MUMBER              188UR TITLE                  TYPE      CATEGORY
  .B-20      STANDARD PROBLEM ANALYSIS              GSI/LI    id B-21      CORE PHYSICS                            GSI/LI    id B-22    LWR YUEL                                GSI        la B-23    LMFBR FUEL                              GSI/LI      ld B-24    SEISMIC QUALIFICATION OF                GSI'      la ELECTRICAL AND MECHANICAL COMPONENTS B-25    PIPING BENCHMARK PROBLEMS                GSI/LI    id B-26    STRUCTURAL INTEGRITY OF                  GSI      '1c CONTAINMENT PENETRATIONS B-27      IMPLEMENTATION AND USE OF              GSI/LI    id SUBSECTION NF B-28      RADIONUCLIDE/ SEDIMENT                  GSI        Ig TRANSPORT PROGRAM B-29      EFFECTIVENESS OF ULTIMATE              GSI/LI-    la HEAT SINKS B-30      DESIGN BASIS FLOODS AND                GSI/LI      id PROBABILITY B-31      DAM FAILURE MODEL                      GSI/RI      id B-32      ICE EFFECTS ON SAFETY                  GSI        la RELATED WATER SUPPLIES B-33      DOSE ASSESSMENT                        GSI/LI    1d METHODOLOGY Amendment F December 15, 1989
 
l' i
CESSAR til&m.
TABLE 11-1 (Cont'd)
I (Sheet 26 of 55)
LISTING OF                                l UNRESOLTED SAFETY ISSUES AMD aEMERIc SAFETY ISSUES ISSUE                                                ISSUE MUMBER                  ' ISSUE TITLE                TYPE.. CATEGORY  i B-34            OCCUPATIONAL RADIATION              GSY      le EXPOSURE REDUCTION B-35            CONFIRMATION OF APPENDIX            GSI/LI    1d        j I MODELS FOR CALCULATIONS                                r OF RELEASES OF                                            ,
RADIOACTIVE MATERIALS IN GASEOUS A i
B-36            DEVELOP DESIGN, TEST,                GSI      2 MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ABSORPTION UNITS.....
B-37            CHEMICAL DISCHARGE TO              -GSI/RI    id RECEIVING WATERS (F)
B-38            RECONNAISSANCE LEVEL                GSI      ig INVESTIGATIONS                                            l B-39            TRANSMISSION LINES                  GSI      lf B-40            EFFECTS OF POWER PIANT              GSI      lf ENTRAINMENT ON PLANKTON B-41            IMPACT ON FISHERIES                  GSI      lf B-42            SOCIOECONOMIC                        GSI      id ENVIRONMENTAL IMPACTS B-43            VALUE OF AERIAL                      GSI      1g PHOTOGRAPHS FOR SITE EVALUATION B-44            FORECASTS OF GENERATING              GSI      1g COSTS OF COAL AND NUCLEAR PLANTS Anendment F December 15, 1989
 
i CESSAR!R Ln.=                                                                                        ;
TABLE A1-1 (Cont'd)
(Sheet 27 of 55)                                                            .,
LISTING QF UMRESOLVED SAFETY ISSUES AMD GENERIC SAFiTY ISSUER t
ISSUE                                                            ISSUE
                                )DlHREE              ISSUE TITLE                                  TYPE _  CATEGORY B-45        NEED FOR POWER - ENERGY                              GSI      le CONSERVATION B-46        COSTS OF ALTERNATIVES IN                              GSI      if ENVIRONMENTAL DESICNS t
B-47        INSERVICE INSPECTION                                  GSI      if                  ;    !
CRITERIA FOR SUPPORT OF CLASS 1, 2, AND 3 AND MC COMPONENTS                                                                              s B-48        BWR CRD MECHANICAL                                    GSI      lb FAILURE (COLLET HOUSING)
B-49        INSERVICE INSPECTION                                  GSI/LI  id CRITERIA AND CORROSION PREVENTION CRITERIA FOR                                                                ,
CONTAINMENTS B-50        POST OPERATING BASIS                                  GSI/RI  ld EARTHQUAKE INSPECTION B-51        ASSESSMENT OF INELASTIC                              GSI      le ANALYSIS TECHNIQUES FOR                                                                ,
EQUIPMENT AND COMPONENTS B-52        FUEL ASSEMBLY SEISMIC AND                            GSI      le LOCA RESPONSES B-53        LOAD BREAK SWITCH                                    GSI/RI  2 B-54        ICE CONDENSER                                        'GSI      lb CONTAINMENTS B-55        IMPROVE RELIABILITY OF                                GSI      lb TARGET ROCK SAFETY RELIEF VALVE (IN BWR'S)
Amendment F December 15, 1989
 
l C E S S A R ti M c.v.                                                                          ;
i T& ELE 11-1 (Cont'd)                                        ,
(sheet 28 of 55)                                          ;
LISTING OF                                              ,
mrnEsoLVED shFETy Issues AND GENERIC smFErr Issues issue                                                      Issue IssDE TITLE                TYPE          CATEGORY            ,
MUMBER B-56                DIESEL GENERATOR                      GSI            2 RELIABILITY t
B-57                STATION BLACKOUT                      GSI              le B-58                PASSIVE MECHANICAL                    GSI              2 i
FAILURES B-59                REVIEW OF (N-1) ICOP                  GSI/RI          id i
OPERATION IN BWRS AND                                                    ,
PWRS      .
B-60                  LOOSE PARTS MONITORING              GSI              2 SYSTEM B-61                  ALLOWABLE ECCS EQUIPMENT            GSI              2
-                                                                                                  i OUTAGE PERIODS B-62                  REEXAMINATION OF                      GSI/LI          id TECHNICAL BASIS FOR-ESTABLISHING SLS AND LSSSS B                  ISOLATION OF LOW PRESSURE            GSI              2                ,
SYSTEM CONNECTED TO THE                                              .j REACTOR COOLANT PRESSURE BOUNDARY B-64                    DECOMMISSIONING OF                  GSI                id REACTORS l
I Amendment F December 15, 1989
 
1 l
1 CESSAR !!!#cui..
l I
l TABLE 11-1 (Cont'd)                                                                                        !
l (Sheet 29 of 55)
LISTING OF                                                                                              l UNEESOLVED SAFETY ISSUES AND GEMIEIC SAFETY ISSUES J
i ISSUB                                                                                                                  ISSUE                            1 3p3333                                      ISSUE TITLE                                                                TYPE            CATEGORY        l l
B-65    IODINE SPIKING                                                                                                GSI              if I      B-66    CONTROL ROOM INFILTRATION                                                                                    GSI              2                l L              MEASUREMENTS                                                                                                                                    l B-67    EFFLUENT AND PROCESS                                                                                          GSI              le MONITORING                                                                                                                                      1 INSTRUMENTATION B-68    PUMP OVERSPEED DURING                                                                                        GSI              If        -
l L              LOCA
'                                                                                                                                                              \
l      B-69    ECC LEAKAGE                                                                                                  GSI              le EX-CONTAINMENT B-70      POWER GRID FREQUENCY                                                                                        GSI              le          i DEGRADATION AND EFFECT 01:                                                                                                                  'l PRIMARY COOLANT PUMPS B    INCIDENT RESPONSE                                                                                            GSI              le l
1      B-72      DEVELOPMENT OF MODELS FOR                                                                                    GSI/LI          1d l              ASSESSING RISK OF HEALTH L              AND LIFE SHORTENING FROM                                                                                                                        '
I URANIUM AND COAL FUEL CYCLE B-73    MONITORING FOR EXCESSIVE                                                                                      GSI              le              j VIBRATION INSIDE THE                                                                                                                            ;
REACTOR VESSEL C-1      ASSURANCE OF CONTINUOUS                                                                                      GSI              2 L                LONG TERM CAPABILITY OF                                                                                                                        l HERMETIC SEALS ON                                                                                                                              l j
INSTRUMENTATION AND ELECTRICAL EQUIPMENT                                                                                                                            l i
Amendment F December 15, 1989
 
i CESSAR Encino.                                                                                                        )
J l
1 l
TABLE A1-1 (Cont'd)                                                                                *
                                                                                                                            ~
(Sheet 30 of $5)
LISTING OF                                                                                    ;
6 S E R__1MD                                                                                            ~
emmERIc saraTY Issues ISSUE                                                                                  ISSUE MUMBER            ISSUI TITLE                                              ,          TYPE      CATEGORY L    C-2      STUDY OF CONTAINMENT                                                          GSI      2
(              DEPRESSURIZATION BY INADVERTENT SPRAY OPERATION C-3      INSULATION USAGE WITHIN+                                                      GSI      le CONTAINMENT C-4      STATISTICAL METHODS FOR                                                        GSI/RI    2 ECCS ANALYSIS C-5      DECAY HEAT UPDATE                            ,
GSI/RI    2          ,
C-6      LOCA HEAT SOURCES                                                              GSI/RI. 2 C-7      PWR SYSTEM PIPING                                                              GSI      Ic C-8      MAIN TEAM LINE ISOLATION                                                      GSI      ib VALVE LEAKAGE CONTROL SYSTEM (IN BWR'S)
C-9      RHR HEAT EXCHANGER TUBE                                                        GSI      If FAILURES C-10    EFFECTIVE OPERATION OF                                                        GSI      2                  4 CONTAINMENT SPRAYS IN A LOCA C-11    ASSESSMENT OF FAILURE AND                                                      GSI      1c RELIABILITY OF PUMPS AND VALVE C-12    PRIMARY SYSTEM VIBRATION                                                      GSI      2 ASSESSMENT C-13    NON-RANDOM FAILURES                                                            CSI      le Amendment F December 15, 1989
 
J CESSAR WWcu..                                                                                            :
i 1
1 TABLE 11-1 (Cont'd)                                                  J (Sheet 31 of 55)
LISTING OF UmmESOLvan SAFETY ISSUES AND                                              i envEmIC SAFETY ISSUES ISSUE                                                                      ISSUE NUMBER                          ISSUE TITLE                                TYPE      CATEGORY C-14        STORM SURGE MODES FOR                                          GSI/LI    if COASTAL SITES C-15        NUREG REPORT FOR LIQUIDS                                      GSI/LI    ld TANK FAILURE ANALYSIS C-16        ASSESSMENT OF                                                  GSI        lg AGRICULTURAL LAND IN RELATION TO POWER PIANT                                                                    !
SITING AND CO6 LING SYSTEM SELECTION C-17        INTERIM ACCEPTANCE                                            GSI        ig CRITERIA FOR                                                                          ~
SOLIDIFICATION AGENTS FOR                                                                  >
RADIOACTIVE SOLID WASTE D-1      ADVISABILITY OF A SEISMIC                                      GSI        la SCRAM D-2      EMERGENCY CORE COOLING                                        GSI        la              [
SYSTEM CAPABILITY'FOR FUTURE PLANTS D-3      CONTROL ROD DROP ACCIDENT                                      GSI        1c HF 1.1    SHIFT STAFFING                                                GSI        id HF 1.2    ENGINEERING EXPERTISE ON-                                      GSI        ld SHIFT                                                                                      -
'HF 1.3        GUIDANCE ON LIMITS AND                                        GSI        id CONDITIONS OF SHIFT WORK MF 1.3.1    HUMAN FACTOR PROGRAM                                          GSI        id PLAN"-TRAINING (F)
Amendment F December 15, 1989
 
4 w '
: q.                                                                                    J
                  +w  -
CESSAR WWicuio,,,                                                                              !,
4 6
I TABLE 11-1 (Cont'd)
(Sheet 32 of 55) i LISTING QF UNRESOLVED SAFETY ISSUES AND                                L GENERIC SAFETY ISSUES
          /
ISSUE                                                        ISSUE
                        ) LUMBER                        ISSUE TITLE                TYPE      CATEGORY
                    -HF 1.3.2                  HUMAN FACTORS. PROGRAM                GSI        id                ;
i                                            - PLAN--LICENSING                                                  1 EXAMINATIONS . (F)                                                ,
f l                      HF 1.3.3'              HUMAN FACTORS PROGRAM                GSI        id-PLAN--PROCEDURES-OPERATING-AND MAINTENANCE (F)                                              ,
HF 1.3.4a              HUMAN FACTORS PROGRAM                GSI        2 PLAN - MAN MACHINE-
* INTERFACE --LOCAL CONTROL
* STATIONS
                                                                                                            ,t
,                  IHF 1.'3.4b-              HUMAN FACTORS PROGRAM                GSI        2                ,
l                                              PLAN --MAN-MACHINE INTERFACE - ANNUNCIATORS ~                                        ;
HF 1.3.4c              HUMAN FACTORS PROGRAM                GSI        2 PLAN'- MAN' MACHINE INTERFACE - OPERATIONAL AIDS' HF.1.3.4d              HUMAN FACTORS PROGRAM                GSI        2 PLAN - MAN MACHINE INTERFACE - AUTOMATION                                          -
AND ARTIFICIAL INTELLIGENCE HF 1.3.4e              HUMAN FACTORS PROGRAM                GSI        2 PLAN - MAN MACHINE INTERFACE'- COMPUTERS AND COMPUTER DISPLAYS-HF 1.3.5              HUMAN FACTORS PROGRAM                GSI        ld PLAN--STAFFING AND                                                -
QUALIFICATIONS (F) 4 Amendment F December 15, 1989
 
CESSAR !nWicciom TABLE 11-1 (Cont'd)
(8heet'33 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY IS8UES
  ,    'i-ISSUE'                                                  ISSUE 6        .
NUNBER                          ISSUE TITLE              TYPE      CATEGORY HF 1.3.6            HUMAN-FACTORS PROGRAM                GSI      id pit.N--MANAGEMENT AND ORGANIZATION (F)                                                    i
                                                                                                    -1
!-              HF 1.3.7            HUMAN FACTORS PROGRAM                GSI.      1d                  l l                                  PLAN--HUMAN PERFORMANCE                                    _. l l                                    (F)                                                        m-    .)
I
(.              .
HF 2.1              EVALUATE INDUSTRY                    GSI      ld    '
TRAINING HF-2.2              EVALUATE INPO                        GSI        id ACCREDITATION HF 2.3              REVISE SRP SECTION 13.2              GSI      id 4
HF 3.1              DEVELOP JOB KNOWLEDGE                GSI      id CATAI4G i
HF 3.2              DEVELOP LICENSE                      GSI        id
: i.                                -EXAMINATION HANDBOOK l-HF 3.3              DEVELOP CRITERIA FOR                GSI        id i
NUCLEAR POWER PLANT SIMULATORS l
HF.3.4              EXAMINATION REQUIREMENT              GSI        id l              HF 3.5              DEVELOP COMPUTERIZED EXAM            GSI        id SYSTEM HF 4.1              INSPECTION PROCEDURE FOR            GSI        id UPGRADED EMERGENCY OPERATING PROCEDURES HF 4.2              PROCEDURES GENERATION                GSI        id PACKAGE EFFECTIVENESS EVALUATION Amendment F December 15, 1989
 
w      -
l '.
CESSARTinkm.                                                        '
TABLE 11-1 (Cont'd)                          +
l (Sheet.34 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND 5                              GENERIC SAFETY ISSUES                        !'
          ' ISSUE ISSUE NUMBER  -
ISSUE TITLE              TYPE    CATEQQBX HF'4,3      CRITERIA FOR                      GSI      id-SAFETY-RELATED OPERATOR
(                        ACTIONS l-HF 4.4      GUIDELINES FOR UPGRADING-        GSI      id OTHER-PROCEDURES HF 5.1      LOCAL CONTROL STATIO!!S          GSI      2 HF:5.2      REVIEW ' CRITERIA FOR HUMAN
* GSI      2 FACTORS ~ ASPECTS OF                                  i
                        -ADVANCED CONTROLS AND INSTRUMENTATION b            HF 5.3      EVALUATION OF OPERATIONAL-        GSI      id AID SYSTEMS HF 5.4      COMPUTERS AND' COMPUTER          GSI      id          ,-
!                        DISPLAYS HF 6.1      DEVELOP REGULATORY                GSI      id POSITION ON MANAGEMENT AND ORGANIZATION HF'6.2-    REGULATORY POSITION ON            GSI      id MANAGEMENT AND L      >
ORGANIZATION AT OPERATING l                        REACTORS HF'7.1      HUMAN ERROR DATA                  GSI      id ACQUISITION HF 7.2      HUMAN ERROR DATA STORAGE          GSI      id AND RETRIEVAL I
HF 7.3      RELIABILITY EVALUATION            GSI      id SPECIALIST AIDS Amendment F December 15, 1989 a  w
 
e; .m.
2 4
CESSAR Gl!Wicui.(                                                                                ;
a'                                      TABLE A1-1 (Cont'd)
                                                    -(Sheet 35 of 55)
LISTING OF UNRESOLVED BAFETY ISSUES AND EENERIC SAFETY ISSUES ISSUE                                                                  ISSUE                    ,
MUMBER                            ISBUE TITLE                          TYPE. g&IEgqBX
            -HF.7.4          SAFETY EVENT' ANALYSIS                                GSI      id            i RESULT APPLICATIONS                                                              i HF 8.0        MAINTENANCE AND                                        GSI      2              '
l                            SURVEILLANCE PROGRAM L
I.A.1.1        OPERATIONAL SAFETY--SHIFT                              GSI      ig TECHNICAL ADVISOR L            I.A.1.2      -OPERATIONAL' SAFETY--SHIFT                              GSI      1g          .:
SUPERVISOR ADMINISTRATIVE                                                      j
                            ' DUTIES.                                                                        .j I.A.1.3        OPERATIONAL SAFETY--SHIFT                              GSI      ig        <
1                            EUMING -
l I.A.1.4      -LONG TERM UPGRADING OF                                  GSI      lg OPERATING PERSONNEL
            'I.A.2.1      .(1-3) IMMEDIATE UPGRADING                              GSI      ig OF OPERATOR AND SENIOR i
OPERATOR TRAINING L-I.A.2.2        TRAINING AND                                          GSI      id            3 QUALIFICATIONS OF OPERATING PERSONNEL I.A.2.3        ADMINISTRATION OF                                      GSI      ig            ^
i                            TRAINING PROGRAMS f
I.A.2.4        NRR PARTICIPATION IN                                  GSI      id INSPECTOR TRAINING I.A.2.5        TRAINING AND                                          GSI      ig QUALIFICATION OF OPERATING PERSONNEL--PLANT DRILLS Amendment F December 15, 1989
 
m 3              2:e                                                                              ,
{                            '
                                                                                                )
            ~
c.
                    'CESSAR mitriemo.                                                            !
                                                                                                )
                                                                                              .1 i
              ,                              TABLE 11 1 (Cont'd).
(8heet 36 of 55)
.y.
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES t.
ISSUE                                            ISSUE-NUMBER                  ISSUE TITLE              TYPE      CATEGORY  .i
:I.A.2.6      (1) IDNG TERM UPGRADING.          GSI      id              l OF TRAINING.
QUALIFICATIONS I.A'.2.6      (2,4,5,6) LONG-TERM-              GSI      ig UPGRADING OF TRAINING AND-                                    i QUALIFICATIONS                                          .
I'A.2.6
                        .          (3) OPERATOR WORKSHOPS            GSI      le              )
I.A.2.7    ACCREDITATION OF= TRAINING-          GSI      ig
                                  . INSTITUTIONS                                                1 L                    I.A.3.1-    REVISE' COPE AND CRITERIA          GSI      ig-FOR LICENSING EXAMS I.A.3.2. OPERATOR LICENSING-                GSI      ig PROGRAM CHANGE l
I-
,                    I.A.3.3      REQUIREMENTS FOR OPERATOR          GSI      le L                                  FITNESS
( ..
I.A.3.4    . LICENSING OF-ADDITIONAL            GSI      ig l                                  OPERATOR PERSONNEL
{
LICENSING OF                        GSI/LI I.A.3.5                                                    id            i PERSONNEL--STATEMENT OF                                      l UNDERSTANDING WITH INPO                                      I l --                              AND DOE I.A.4.1      (1-2) TRAINING SIMULATOR          GSI      lg IMPROVEMENT--INITIAL l'
I.A.4.2      (1) TRAINING SIMULATOR            GSI/TMI  le IMPROVEMENTS--LONG TERM I.A.4.2      (2) TRAINING SIMULATOR            GSI/TMI  ig IMPROVEMENTS -- LONG TERM-Amendment F December 15, 1989 L
 
        ~[.            t
  +!.
CESSAR1!innean.                                                                    '
1 l
l I
l l
TABLE Al-l (Cont'd)                                  -1 (sheet 37 of 55)
LISTING OF UNRESOLVED SAFETY IS8UES AND GENERIC SAFETY ISSUES issue                                              IsscE-                          ;
NUMBER                  ISSUE TITLE              TYPE      CATEGORY
                                                                                                  ]
                                  .    -                                                        1 i
I.A.4.2      (3) TRAINING SIMULATOR              GSI/TMI  id IMPROVEMENTS -- LONG TERM.
I.A.4.2      (4) TRAINING SIMULATOR              GSI/TMI  le IMPROVEMENTS -- LONG TERM I'.A.4.3-  - FEASIBILITY STUDY FOR-              GSI/LI    id        /
l                          PROCUREMENT.OF. TRAINING                                            ,
l=                        SIMULATOR                                                            1 I.A.4.4      FEASIBILITY STUDY TO                  GSI/LI    id        ...
EVALUATE POTENTIAL VALUE                                            '
OF NRC ENGINEERING COMPUTER                                                      ,
I.B.1.1        (1-4)' ORGANIZATION AND            GSI/TMI  le
                          . MANAGEMENT - LONG TERM                                    -
IMPROVEMENTS I.B.1.1        (5) MANAGEMENT FOR                  GSI      ig                -t OPERATION--LONG-TERM                                              1 IMPROVEMENTS (F)
I.B.1.1        (6&7) ORGANIZATION AND              GSI/TMI  le MANAGEMENT - LONG TERM IMPROVEMENTS I.B.1.2        (1-3) MANAGEMENT FOR                GSI/LI  lg OPERATIONS--EVALUATION OF NTOL APPLICANTS r
l:            I.B.1.3        (1-3) MANAGEMENT FOR                GSI/LI  id OPERATIONS--LOSS OF SAFETY FUNCTION I.B.2.1        (1-7) REVISION OF IE                GSI/LI  id INSPECTION PROGRAM Amendment F December 15, 1989
 
CESSAR ;!!Lu..
I TABLE A1-1 (Cont'd)                                  l 1
(sheet 3a of 55)
LISTING OF                                      '
UNRESOLVED SAFETY ISSUES AND                              l GENERIC SAFETY ISSUES                                ,
issue                                            issue NUMBER            ISSUE TITLE                    TYPE      CATEGORY l
  ~I.B.2.2 RESIDENT INSPECTORS AT                    GSI/LI    ig        -)
OPERATING REACTORS I.B.2.3 INSPECTIONS AT-OPERATING                  GSI/LI    id          I REACTORS--REGIONAL                                              l EVALUATIONS-                                                  j 1
;  I.B.2.4 OVERVIEW OF LICENSEE                      GSI/LI    id l'          PERFORMANCE
  -I.C.1    (1-4)-SHORT TERM ACCIDENT                GSI      2 ANALYSIS AND PROCEDURES                                        .
REVISION                                                        )
  .I.C.2  SHIFT AND RELIEF TURNOVER                GSI      ig PROCEDURES I.C.3  SHIFT SUPERVISOR                          GSI      ig RESPONSIBILITIES                                              j I.C.4  OPERATING                                GSI      ig PROCEDURES--CONTROL ROOM ACCESS I.C.5  PROCEDURES'FOR FEEDBACK                  GSI/TMI  ig OF OPERATING EXPERIENCE I.C.6  PROCEDURE FOR                            GSI      id VERIFICATION OF CORRECT                                          l PERFORMANCE OF OPERATING                                        :
ACTIVITIES l
I.C.7  NSSS VENDOR REVIEW OF                    GSI      ig          !
OPERATING PROCEDURES Amendment F December 15, 1989
 
_. ,y ,,                                                                    .,
                                    . .  .        .~      .-. _.  .      ._.
                                                                                              \
CESSAR E!Miccior                                                      l f
TABLE A1-1 (Cont'd)
(Sheet 39 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE                                        ISSUE
    . g.,            NUMBES            ISSUE TITLE                TYPE        CATEGORY
                    'I.C.8  PII4T MONITORING OF                  GSI          ig ELECTED EMERGENCY                                              ,
                            -PROCEDURES FOR NTOL                                          'l APPLICANTS l
I.C.9  LONG TERM PLAN FOR                    GSI          le UPGRADING OF PROCEDURES I.D.1  CONTROL ROOM DESIGN                  GSI/TMI      id REVIEWS -- GUIDELINES AND REQUIREMENTS
,                    I.D.2  -CONTROL ROOM DESIGN                  GSI/TMI- 2
: l.                            REVIEWS -- PLANT SAFETY                                        .
PARAMETER DISPLAY CONSOLE                                    :4 l-                    I.D.3  CONTROL ROOM DESIGN --                GSI/TMI      2-SAFETY SYSTEM STATUS                                          4 i                            MONITORING L
f                    I.D.4  CONTROL ROOM DESIGN-                  GSI          2 STANDARD-                                                      ;
I.D.5    (1) CONTROL ROOM DESIGN              GSI          2 i.
i                            -- IMPROVED l-                            INSTRUMENTATION RESEARCH ALARMS AND DISPLAYS l                    I.D.5    (2) CONTROL ROOM DESIGN              GSI          2 l'                            -- IMPROVED L..
INSTRUMENTATION RESEARCH I.D.5    (3) CONTROL ROOM DESIGN              GSI/LI      2 l:                            -- ON-LINE REACTOR SURVEILLANCE SYSTEMS                                            1 l
I I
Amendment F December 15, 1989
 
CESSAR !Ene.m.                                                                  1 l
l l
TABLE Al-1 (Cont'd) 1 (sheet 40 of 55)                                J LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUE 8 ISSUE                                                  ISSUE NUMBER                        ISSUE TITLE              TYPE.. CATEGORY I.D.5          (4) CONTROL ROOM DESIGN                GSI      2
                    - IMPROVED t
INSTRUMENTATION RESEARCH I.D.5          (5) DISTURBANCE ANALYSIS              GSI/LI    le SYSTEMS-l  I . D. 6 '  ' CONTROL ROOM                            GSI/LI    id-DESIGN--TECHNOLOGY TRANSFER CONFERENCE I.E.1-        ESTABLISH OFFICE FOR                    GSI/LI    id ANALYSIS AND EVALUATION OF' OPERATIONAL DATA I.E.2        PROGRAM                                GSI/LI    id OFFICE--OPERATIONAL DATA EVALUATION t
I.E.3-        OPERATIONAL SAFETY DATA-                GSI/LI    id
;                - ANALYSIS I.E.4          COORDINATION OF LICENSEE,              GSI/LI    id INDUSTRY, AND REGULATORY PROGRAMS I.E.5          NUCLEAR PLANT RELIABILITY              GSI/LI    id DATA SYSTEM                                                    ,
I.E.6          REPORTING REQUIREMENTS                GSI/LI    id FOR REACTOR OPERATING EXPERIENCE I.E.7          INFORMATION FOR ANALYSIS                GSI/LI  id AND DISSEMINATION OF OPERATING EXPERIENCE--
FOREIGN SOURCES l                                                            Amendment F December 15, 1989
 
1 CESSAR1H Lu.*
TABLE A1-1 (Cont'd)
(Sheet 41 of 55)
LISTING OF UMRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE                                      ISSUE NUMBER            ISSUE TITLE              TYPE. CATEGORY I.E.8    HUMAN ERROR RATE ANALYSIS          GSI/LI    ld I.F.1    QUALITY ASSURANCE -                GSI      2 EXPAND QUALITY ASSURANCE LIT FOR EQUIPMENT IMPORTANT TO SAFETY
                        ~
I.F.2    (1,4,5,7,8,10,11) QUALITY          GSI/TMI  la ASSURANCE--DEVELOP MORE DETAILED QA CRITERIA    ,
I.F.2    (2)-QUALITY ASSURANCE --          GSI/TMI  2 DEVELOP MORE DETAILED CRITERIA I.F.2    (3) QUALITY ASSURANCE --          GSI/TMI  2 DEVELOP MORE DETAILED-CRITERIA I.F.2-  (6)-QUALITY ASSURANCE --          GSI      2 DEVEIDP MORE DETAILED QA CRITERIA I.F.2    (9) QUALITY ASSURANCE --          GSI      2 DEVELOP MORE DETAILED QA CRITERIA I.G.1  SCOPE OF TEST                      GSI        ig PROGRAM--PREOPERATIONAL AND LOW POWER TESTING I.G.2  SCOPE OF TEST PROGRAM              GSI        19 II.A.1  SITING POLICY                      GSI      1c REFORMULATION II.A.2  SITE EVALUATIONS OF                GSI      le EXISTING FACILITIES Amendment F December 15, 1989
 
  ~ CESSAR nascam,.
TABLE 11-1 (Cont'd)
                          -(Sheet 42 of 55)
LISTING OF UNkE80LVED SAFETY ISSUE 8 AND GENERIC 8AFETY 188UE8 ISSUE                                        ISSUE NUMBER            ISSUE TITLE                TYPE      CATEGORY II.B.1  SAFETY REVIEW                        GSI/TMI  2 CONSIDERATION - -REACTOR COOLANT SYSTEM VENTS II.B.2  SAFETY REVIEW                        GSI/TMI  2 CONSIDERATION -- PLANT
~'
SHIELDING TO PROVIDE POST ACCIDENT ACCESS TO VITAL
            ~ AREAS II.B.3  SAFETY REVIEW                        GSI/TMI  2 CONSIDERATION -- POST ACCIDENT SAMPLING SYSTEM II.B.4  SAFETY REVIEW                        GSI-      lg CONSIDERATION--TRAINING TO MITIGATE CORE DAMAGE II.B.5  (1&2) BEHAVIOR OF                  GSI/LI    id SEVERELY DAMAGED FUEL &
CORE MELT II.B.5  (3) EFFECT OF H2 BURNING          .GSI/LI    id AND EXPLOSIONS ON CONTAINMENT STRUCTURE II.B.6  RISK REDUCTION FOR                  GSI      id OPERATING. REACTORS WITH SITES WITH HIGH POPULATION DENSITIES II.B.7  SAFETY REVIEW                        GSI      id CONSIDERATION -- ANALYSIS OF HYDROGEN CONTROL Amendment F December 15, 1989
 
C E S S A R ii!Mne m .
t TABLE 11-1 (Cont'd).                                                                ,
(Sheet 43 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND 4
GENERIC SAFETY ISSUES ISWJE                                                                              ISSDE                                .
NUMBER                              ISSUE TITLE                                    TYPE        CATEGORY II.B.8                    RULEMAKING PROCEEDINGS ON                                GSI          id DEGRADED CORE ACCIDENT-HYDROGEN RULE, SEVERE ACCIDENT, ETC.
II.C.1                    INTERIM RELIABILITY                                    GSI          1c EVALUATION PROGRAM                                                                        y.
II.C.2                    CONTINUATION OF INTERIM                                GSI          1c RELIABILITY EVALUATION
* PROGRAM II.C.3                    RISK ASSESSMENT--SYSTEMS                                GSI          le INTERACTION II.C.4                    RELIABILITY ENGINEERING                                GSI/TMI      2 1
II.D.1                    COOLANT SYSTEM VALVES --                                GSI/TMI      2 TESTING REQUIREMENTS l
I      II.D.2                    COOLANT SYSTEM                                          GSI/TMI      la l                                VALVES--RESEARCH ON TEST                                                                        ,
L                                  REQUIREMENTS l-l'      II.D.3                    COOLANT. SYSTEM VALVES --                              GSI/TMI      2 L                                VALVE POSITION INDICATION II.E.1.1                  AUXILIARY FEEDWATER                                      GSI/TMI      2 SYSTEM EVALUATION I      II.E.1.2                  AUXILIARY FEEDWATER                                      GSI/TMI      2 l                                  SYSTEM AUTOMATIC
!                                  INITIATION AND FLOW L                                  INDICATION l
l' Amendment F December 15, 1989 i
 
CESSARi!,*ainem ,,
~
TABLE 11-1 (Cont'd)
(Sheet 44 of 55)
LISTIn}G OF UNR3 SOLVED SAFETY ISSUES AND g3NERIC SAFETY ISSUES Issun                                                Issus NUMBER                  ISSUE TITLE                  TYPE        CATEGORY GSI/LI II.E.1.3        UPDATE STANDARD REVIEW                            id PLAN AND DEVELOP REGULATORY GUIDE II.E.2.1        RELIANCE ON EMERGENCY                GSI          le CORE COOLING SYSTEM
      , II.E.2.2        RESEARCH ON SMALL BREAK              GSI          1c LOCAs AND ANO!!ALOUS TRANSIENTS        -
II.E.2.3      . UNCERTAINTIES IN ECCS                GSI/TMI      la PERFORMANCE PREDICTIONS i;      II . E. 3'.'1    DECAY HEAT REMOVAL --                GSI/TMI      le RELIABILITY OF POWER-SUPPLIES FOR NATURAL CIRCULATION II.E.3.2        DECAY HEAT.                          GSI          le REMOVAL--SYSTEMS RELIABILITY II.E.3.3        COORDINATED STUDY OF                GSI          le SHUTDOWN HEAT REMOVAL l                      REQUIREMENTS II.E.3.4        DECAY HEAT REMOVAL --                GSI          id ALTERNATE CONCEPTS RESEARCH II.E.3.5        DECAY HEAT                          GSI          le REMOVAL--REGULATORY GUIDE l
: i.      II.E.4.1        CONTAINMENT DESIGN --                GSI          2 p                        DEDICATED PENETRATIONS l'
l I
l Amendment F December 15, 1989
 
                        -. _ -              -      - -    ~~ .      .-    ~ .-
1 LCESSAR1!n%mo,,                                                                    )
                                                                                      \
TABLE Al-1 (Cont'd)                                  !
(Sheet 45 of 55)
LISTING OF UMRE80LVED SAFETY ISSUES AND GENERIC SAFETY IS8UES                                ;
18803                                              ISSUE                        ,
MUMBER                    ISSUE TITLE              TYPE      CATEGORY            ,
1 II.E.4.2    CONTAIN5EtiT DESIGN --                  GSI/TMI  2 h
ISOLATION DEPENDABILITY L
II.E.4.3    CONTAINMENT INTEGRITY.                  GSI/TMI. id CHECK-II.E.4.4    (1-5) CONTAINMENT DESIGN              GSI/TMI  2              E
                  -- PURGING II.E.5.1  B&W DESIGN EVALUATION                  GSI      lb
    .II.E.5.2*  B&W REACTOR TRANSIENT                  GSI      lb RESPONSE TASK FORCE
!    II.E.6.1  TEST ADEQUACY STUDY                    GSI      id
;,                (VALVE, HANGERS, ETC.)
II.F.1    ADDITIONAL ACCIDENT                    GSI/TMI  2 MONITORING j'                INSTRUMENTATION II.F.2      IDENTIFICATION AND                    GSI/TMI  2 RECOVERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING II.F.3      INSTRUMENTATION FOR                    GSI/TMI  2 MONITORING ACCIDENT CONDITIONS II.F.4      STUDY OF CONTROL AND                  GSI/TMI  if PROTECTION ACTION DESIGN i-l    II.F.5    CLASSIFICATION OF I & C,                GSI/LI    id AND ELECTRICAL EQUIPMENT Amendment F December 15, 1989 l
 
e          .      ,
q CESSAR1!nL mt                                                                                                                            j TABLE A1-1 (cont'd)                                                                            >
(Sheet 46 of 55) 1 W TING OF cxRESoLVED SAFETY ISSUES AND GEMERIC SAFETY ISSUES
          . ISSUE                                                                                            ISSUE NUMBER                                    -ISSUE TITLE-                                            TYPE        CATEGORY II.G.1                          POWER SUPPLIES-FOR                                                GSI/TMI    2
                                          . PRESSURIZER RELIEF
!~                                          VALVES, LOCK VALVES, AND LEVEL' INDICATORS
          ;II.H.11                      . MAINTAIN SAFETY OF TMI                                        GSI        ig AND MINIMIZE ENVIRONMENTAL IMPACT-                                                                                    ,
          .II.H.2  .                      OBTAIN DATA ON INSIDE                                            GSI/TMI    ig
                                          . CONDITION OF TMI.
CONTAINMENT II.H.3                        . EVALUATE AND FEEDBACK                                            GSI        le INFORMATION OBTAINED FROM TMI II.H.4                          DETERMINE IMPACT OF TMI                                          . GSI/ LI    id                          .
ON SOCIOECONOMIC AND REAL.
!                                          PROPERTY VALUES
(
II.J.1.1                        ESTABLISH A PRIORITY                                              GSI/LI      id SYSTEM FOR CONDUCTING                                                                                  .
                                          -VENDOR INSPECTIONS l'          II.J.1.2                        MODIFY EXISTING VENDOR                                            GSI/LI      1d l                                          INSPECTION PROGRAMS                                                                                          I i
II.J.1.3                        INCREASE REGULATORY                                              GSI/LI      id                            !
L,                                          CONTROL OVER PRESENT NON                                                                                  :
1 LICENSERS                                                                                              1 II.J.1.4                        ASSIGN RESIDENT                                                  GSI/LI      id INSPECTORS TO REACTOR VENDORS AND ARCHITECT-ENGINEERS I
Amendment F                            )
December 15, 1989                      '
H    -- -      -      _ _ . _ . . _ . .          . _ _ . _ _ _ _ . _ _ _        __ _ _ _ _ _ _ _ . _ _ _          ._      _  _ . . _ . . . _
 
4    CESSAR !!n%mo.
TABLE A1-1 (Cont'd)
(Sheet 47 of 55)
LI8 TING OF UMRESOLVED SAFETY ISBUES AND GENERIC 8AFETY ISSUES IS8UE                                      ISSUE MUMBER              ISSUE TITLE            TYPE.      CATEGORY II.J.2.1  REORIENT CONSTRUCTION            GSI/LI    id INSPECTION PROGRAM II.J.2.2  INCREASE EMPHASIS,0N              GSI/LI    id INDEPENDENT MEASUREMENT IN CONSTRUCTION                                          ..
INSPECTION PROGRAM                                      'j II.J.2.3 ASSIGN RESIDENT                    GSI/LI    ld INSPECTORS TO ALL CONSTRUCTION SITES II.J.3.1  ORGANIZATION AND STAFFING        GSI        le TO OVERSEE DESIGN AND CONSTRUCTION II.J.3.2  MANAGEMENT FOR. DESIGN AND        GSI        le CONSTRUCTION--ISSUE REG.
GUIDE II.J.4.1  REVISE DEFICIENCY REPORT          GSI        id REQUIREMENTS II.K.1    (1,2,4 (a-c) ,5,7,8,10-13,      GSI        i 17-23) MEASURES TO MITIGATE SMALL BREAK LOCA'S AND FEEDWATER ACCIDENTS--IE BULLETINS II.K.1    (3,4d,6,9,14-16,24-28)          GSI        2 MEASURES TO MITIGATE SMALL BREAK LOCA'S & LOSS OF FW ACCIDENTS IE BULLETINS Amendment F December 15, 1989
 
                                          .          .    .        . .-                                                            - ~. - . - - -
.,g'
;                                                                                                                                                        j i
CESSAR !!nincam,.                                                                                                                  I I
1 TABLE Al-1 (Cont'd)
(Sheet 48 of-55)
LISTING OF                                                                                        i l
UNEESOLVED SAFETY ISSUES AND SEMIBIELAAHZI._URBA l
              . ISSUE                                                                          ISSUE                                                ,
MUMBER                  ISSUE TITLE                                            TYPE                                  CATEGORY II.K.2        (1-15,18,20,21)                                                  GSI                                    lb COMMISSION ORDERS ON B&W PLANT TO MITIGATE
                                ' ACCIDENTS' II.K.2'      (16, 17, 19) COMMISSION                                          GSI/TMI                                lb        -t ORDER ON B&W PLANTS TO MITIGATE ACCIDENTS II.K.3        (1-4,7,9-24,26-29,32-54',56,
* GSI                                    1
: 57) FINAL
                                  ^ RECOMMENDATIONS'OF B&O.                                                                                            ,
                                -TASK FORCE TO MITIGATE ACCIDENTS II.K.3      (5,6,8,25,30,31,55)                                              GSI/TMI                                2 FINAL RECOMMENDATIONS OF B&O TASK FORCE TO MITIGATE ACCIDENTS III.A.1.1    (1-2) UPGRADE EMERGENCY                                          GSI                                    ig        l PREPAREDNESS III.A.1.2    (1-3)-UPGRADE LICENSEE                                          GSI/LI                                  id EMERGENCY SUPPORT FACILITIES III.A.1.3    (1) MAINTAIN SUPPLIES OF                                        GSI                                    ig THYROID LOCKING AGENT III.A.1.3-    (2) MAINTAIN SUPPLIES OF                                          GSI                                    ig THYROID BLOCKING AGENT l
Amendment F                              '
I December 15, 1989
 
      '6 ' <
CESSAR !!nificuio                                                                                            l 1
TABLE A1-1 (Cont'd)                                                                ,
                                        .(Sheet 49 of 55)L LISTING OF UNRESOLVED SAFETY ISSUES AND GEEERIC SAFETY ISSUES
      ,      ISSUE'                                                      ISSUE                                          '
NUMEER                  -
ISSUE' TITLE                        TYPE          CATEGORY III.A.2.1      (1-4) AMENDMENT TO                          GSI          id L.
10 CFR 50 AND APPENDIX E
[
            .III.A.2.2-    DEVEIDPMENT OF GUIDANCE                      GSI          ig                                ,
AND CRITERIA GSI/LI l            III.A.3.1      (1-5) EMERGENCY                                            id                      -
PREPAREDNESS -- NRC ROLE-                                                                      ,
IN RESPONDING TO NUCLEAR                        *                                            '
EMERGENCIES III.A 3.2      EMERGENCY PREPAREDNESS --                    GSI/LI.      1d                                <
                            -IMPROVE-OPERATIONS CENTERS u
III.A.3',3      (1-2) EMERGENCY                              GSI/LI          d PREPAREDNESS --                                                                              ,
I-COMMUNICATIONS III.A.3.4      NUCLEAR DATA LINK                            GSI          id                                i III.A.3.5      EMERGENCY PREPAREDNESS --                    GSI/LI        id TRAINING, DRIILS & TESTS III.A.3.6      (1-3) EMERGENCY                              GSI/LI        id PREPAREDNESS -- NRC AND OTHER AGENCIES III.B.1      TRANSFER OF EMERGENCY                          GSI/LI      id PREPAREDNESS RESPONSIBILITIES TO FEMA III.B.2        (1-2) IMPLEMENTATION OF                      GSI/LI      id NRC'S AND FEMA'S RESPONSIBILITIES l~
Amendment F December 15, 1989 l
 
g.
LCESSAR 12%u. -
TABLE 11-1 (Cont'd)
(Sheet 50 of 55)
LISTING OF WRE80LVED SAFETY ISSUE 8 AND
  '"                              GENERIC BAFETY ISSUES
          . ISSUE                                          ISSUE gggggg                ISSUE TITLE              TYPE        CATEGORY III.C.1    (1-3) PUBLIC                      GSI/LI      id INFORMATION--PROVIDE TO NEWS MEDIA AND PUBLIC III.C.2    (1-2) PUBLIC                      GSI/LI      id INFORMATION--PROVIDE TRAINING III.D.1.1  (1) PRIMARY COCIMIT                GSI/TMI    if SOURCES OUTSIDE THE
* CONTAINMENT STRUCTURE'
            -III.D.1.1    (2)_ REVIEW.INFORMATION ON        GSI          la PROVISIONS FOR LEAK DETEC* ION III.D.1.1  (3) DEVELOP PROPOSED              GSI          la SYSTEM ACCEPTANCE CRITERIA III.D.1.2  RADIOACTIVE GAS                    GSI/TMI      1F MANAGEMENT III.D.1.3    (1-3) VENTILATION SYSTEM          GSI/TMI      lf AND RADICIODINE ABSORBER CRITERIA III.D.1.3    (4) VENTILATION SYSTEM            GSI          id AND RADICIODINE. ABSORBER CRITERIA III.D.1.4  RADWASTE SYSTEM FEATURES            GSI/TMI    lf TO AID IN ACCIDENT III.D.2.1    (1-3) RADIOLOGICAL                  GSI/Th'!I  la MONITORING OF EFFLUENTS Amendment F December 15, 1989
 
I 5 I. -
LCESSARlinhn.                                                                      i TABLE 11-1 (Cont'd)
(Sheet 51 of 55)
LISTING OF UNRESOLVED 8AFETY IS8UES AND GENERIC BAFETY ISBUE8 l
                                                                                            .q I8 SUE                                                ISSUE MUMBER                    ISBUE TITLE                TYPE      CATEGORY
          .III.D.2.2      (1) RADICIODINE,                      GSI      id                l CARBON-14, AND-TRITIUM                                            'l PATHWAY DOSE ANALYSIS                                                l l
III.D.2.2      (2-4) RADIOIODINE,                    GSI      le                !
CARBON-14, AND TRITIUM                                      . ,.
PATHWAY DOSE' ANALYSIS III.D.2.3      (1)' LIQUID PATHWAY                  GSI      1c RADIOLOGICAL CCNTROL III.D.2.3    '(2)-SCREENING OF SITES                GSI      ic FOR LIQUID PATHWAY
  !          .            CONSEQUENCE III.D.2.3      (3)-LIQUID PATHWAY                    GSI-      ic INTERDICTION-l III.D.2.3      (4)
 
==SUMMARY==
ASSESSMENT OF            GSI      1c LIQUID-PATHWAY CONSEQUENCES III.D.2.4      (1) OFFSITE DOSE                      GSI      ld MEASUREMENTS III.D.2.4      (2) OFFSITE DOSE                      GSI/LI    1d MEASUREMENTS III.D.2.5-    OFFSITE DOSE                            GSI/LI    1c CALCULATIONAL MANUAL III.D.2.6      INDEPENDENT RADIOLOGICAL              GSI/LI    id MEASUREMENTS III.D.3.1    RADIATION PROTECTION                    GSI      1c PLAN Amendment F December 15, 1989
 
f
          '.I'-                                                                                                                                                          T
    , ;1                                                                                                                                                            -.
CESSAR !!nhu
* j if TABLE A1-1 (Cont'd)
                                                    '(Sheet 52 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE-                                                                            ISSUE                                                        ''-
NUMBER                          ISSUE TITLE                                        ZUj._            CATEGORY i
III.D.3.2            (1-4) HEALTH PHYSICS                                        GSI/LI          id.
IMPROVEMENTS                                                                                                                    _;
III.D.3.3            (1-4)-IN-PLANT RADIATION                                    GSI/TMI          2 L                                      MONITORING III.D.3.4.          CONTROL ROOM HABITABILITY                                    GSI              2 III.D.3.5"            (1-3). RADIATION WORKER                                      GSI/LI          id                                                ,
EXPOSURE DATA BASE.                          ,
IV.Al                SEEK LEGISLATIVE                                              GSI/LI          id AUTHORITY IN ENFORCEMENT PROCESS IV.A.2              REVISE ~ ENFORCEMENT POLICY                                  GSI/LI          ld IV.B.1'              REVISE PRACTICES F0F.                                        GSI/LI          ld                                                ,
ISSUANCE OF INSTRUCTIONS                                                                                                        -
AND~INFORMATION TO LICENSEES IV.C.1              EXTEND LESSONS. LEARNED                                      GSI/LI          id FROM TMI TO OTHER NRC PROGRAMS IV.D.1              NRC STAFF TRAINING                                            GSI/LI          ld IV.E.1              EXPAND RESEARCH ON                                            GSI/LI          id l                                      QUANTIFICATION OF SAFETY DECISION-MAKING IV.E.2              PLAN FOR EARLY RESOLUTION                                    GSI/LI          id OF SAFETY ISSUES Amendment F December 15, 1989
                    .        . . ~ . .            . - . . . . . _ - - - - _ - _ - -  , _ _ . . _ -
 
g, r,
1 o!
e'-        CESSARnnLmt TABLE 11-1 (Cont'd)
(Sheet 53 of 55)-
LISTING OF
  -                        UMRESOLVED SAFETY 188UE8 AND G3MERIC SAFETY ISSUES ISSUE                                      ISSUE MUMBER            ISSUE TITLE            TYPE      CATEGORY IV.E.3    PLAN FOR RESOLVING ISSUES        GSI/LI    id AT THE CONSTRUC7' ION PERMIT STAGE IV.E.4  RESOLVE GENERIC ISSUES BY        GSI/LI    1d RULEMAKING IV.E.5-  ASSESS CURRENTLY                  GSI        1c OPERATING PLANTS                            ,
GSI/LI
            -IV.F.1    INCREASE IE SCRUTINY OF                      Id POWER ASCENSION TEST                                    ^
PROGRAM
    -,-    -IV.F.2-  -EVALUATE THE IMPACTS OF          GSI/LI      ld FINANCIAL DISINCENTIVES
                      'TO--THE SAFETY OF NUCLEAR POWER PLANTS IV.G.1  DEVELOP A PUBLIC-AGENDA          GSI/LI      id FOR RULEMAKING IV.G.2  PERIODIC AND SYSTEMATIC          GSI/LI      id REEVALUATION OF EXISTING RULES IV.G.3  IMPROVE RULEMAKING              GSI/LI      id PROCEDUKES IV.G.4  STUDY ALTERNATIVES FOR            GSI/LI      id IMPROVED RULEMAKING PROCESS IV.H.1  NRC PARTICIPATION IN THE          GSI/LI      id RADIATION POLICY COUNCIL l
Amendment F December 15, 1989 e
: q.                                                                                              J 5
              -CESSAR liniNma.                                                                              I l
TABLE A1-1 (Cont'd)                                                    l 1
(Sheet 54 of 55)                                                      .
LISTING OF unmEsOLvEn axFETr Issues xwn GEMERIC SAFETY ISSUES ISSUE                                        ISSUE M                  issue TITLE              ZIEE._            CATEGORY V.A.1. DEVELOP NRC POLICY                GSI/LI            id STATEMENT ON SAFETY V.B.1    STUDY AND RECOMMEND, AS.          GSI/L1            Id APPROPRIATE, ELIMINATION-OF NONSAFETY RESPONSIBILITIES V.C.1    STRENGTHEN THE ROLE OF            GSI/LI            id ADVISORY COMMITTEE ON REACTOR-SAFEGUARDS 9
V.C.2    STUDY NEED FOR ADDITIONAL          GSI/LI            id                          ,
ADVISORY COMMITTEES l-D              V.C.3    STUDY THE NEED TO                  GSI/LI            id
                        . ESTABLISH AN-INDEPENDENT NUCLEAR SAFETY BOARD V.D.1    . IMPROVE PUBLIC AND                GSI/LI            id                          j INTERVENOR PARTICIPATION-IN THE HEARING PROCESS
!              V.D.2    STUDY                              GSI/LI            1d CONSTRUCTION-DURING-ADJUDICATION RULE
              -V.D.3    REEXAMINE COMMISSION ROLE          GSI/LI              ld IN ADJUDICATION i
V.D.4      STUDY THE REFORM OF THE          GSI/LI              id                          ,
LICENSING PROCESS V.E.1    STUDY THE NEED FOR                GSI/LI              id TMI-RELATED LEGISLATION Amendment F December 15, 1989 i-I
                                                                    , . . , . . . _ . _ ~ . . , . . ,    ,a
 
e                            .                                                                                                                        .
      . h
      's          p LCESSAR M Lu.
                                                                                                                                                      ~ .
TABLE 11-1~(Cont'd)                                                                                            ,
(8heet 55 of 55)
LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC 8AFETY ISSUES                                                                                          i 7
                                  '    IS8UE                                                                                                                    IS8UE MUMBER                                                        TSSUE TITLE                                                TYPE          CATEGORY y                                    V . F.1 -                                            STUDY NRC TOP MANAGEMENT.                                            GSI/LI        id                                  ,
STRUCTURE AND PROCESS V.F.2                                                REEXhMINE ORGANIZATION                                              GSI/LI        id
                                                                                          .AND FUNCTIONS OF THE NRC OFFICE                                                                                                      -
3 V'F.3 .                                              REVISE DELEGATION OF                                                GSI/LI        id AUTHORITY TO STAFF
                                  ' V.F.4:                                                  CLARIFY AND STRENGTHEN _                                            .GSI/LI        id
                                                                                          .THE RESPECTIVE ROLES OF                                            *                                          '
                                                                                        ' CHAIRMAN, COMMISSION, AND                                                                                                r EXECUTIVE DIRECTOR FOR.
OPERATIONS
                                                                                                                                                                                                          '~
l V.F. 5'                                            . AUTHORITY TO DELEGATE                                              GSI/LI=        id EMERGENCY RESPONSE FUNCTIONS TO A' SINGLE COMMISSIONER
                          ,        - V.G.1                                                ACHIIVE SINGLE LOCATION,'                                            GSI/LI        id                                ,
LONG-TERM.                                                                                                            ,
I lI                                                                                                                                                                                                                '
V.G.21                                              ACHIEVE. SINGLE IDCATION,                                            GSI/LI        id INTERIM i
i ,
ci ,
Amendment F December 15, 1989 l
              -    4.---  .-_  ___m  - _ _ _ _ _ _ _ . _ _ _ _    _._ __._.__ ____ ___              _ _ _ _ _ _ _    _  _ _ _ .- . _ , . . _ - . . ,  - -
                                                                                                                                                                  ,.#.,.        --w..- - - -, = , - . , . , . . -
 
m
                        ;M> s CESSAR1!n h o.
2.0'        LIST Olf UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN Table A2-2 of this section identifies the USIs,- and Medium- and High-priority GSIs which are - technically relevant to-the System-      ,
80+ Standard Design, consistent with 10 CFR Part 52.47.            The  j
  '7'6 ,
process for - identification of these issues is provided in the Overview and in Section 1.0 of this appendix,            along with a definition of the " Issue Types" indicated within the list given          ,
in this section.                                                        l i
i l
l' L.
1 l
l l
\
l L                                                                      Amendment F A-3                Decerter 15, 1989
 
  .                    m,    ..                                                                                                                                                    .
i t-  .c t b
1(
LCESSAR1!nhion t
f i
  !- l e
y t.
4 THIS PAGE INTENTIONALLY BLANK:
1 4
4 4
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11 Amendment F A-4                                                  December 15, 1989
                          .a
                                              - , , , - , . . . . . _ . .      . _ _ _ _ . - ,~_,. .... ..._.,.,_,..._., ..,_. , ,          ,  . . . . _ . . . . . . _  . . . ,
 
                                                                      ,                                                                                                                    1 l
l CESSAR1!n% mot                                                                                                                                                                  J
                                                                                                                                                                                        'l l
TABLE A2-1 (Sheet l'of 10)                                                                                                          ,
I' LIST OF UNRESOLVED SAFETY ISSUES AND HIGE/ MEDIUM                                                                                                              "
PRIORITY GENERIC ISSUES APPLICABLE TO THE SYSTEM 30+ STANDARD DESIGN                                                                                                        'j
                                                                                                                                                                                        -I
                                                                                                                                                                                          ~
ISSUE                                                                                                                                                            ISSUE    -)
NUMBER                                                ISSUE TITLE
* TYPE          ,
3                              SETPOINT DRIFT IN                                                                                                              GSI INSTRUMENTATION                                                                                                                              '
14                            PWR PIPE CRACKS                                                                                                                GSI        .j 15                          -RADIATION-EFFECTS ON REACTOR                                                                                                    GSI            i VESSEL SUPPORTS                                                                                                                              i 22                            INADVERTENT BORON' DILUTION                                                                                                    GSI EVENTS J
            - 23                            REACTOR' COOLANT PUMP SEAL                                                                                    ,                GSI            ,
L FAILURES l:                                                                                                                                                                          GSI 29                            BOLTING DEGRADATION OR FAILURES IN NUCLEAR PLANTS 36                            LOSS OF SERVICE WATER                                                                                                          GSI            ,
45                            INOPERABILITY OF INSTRUMENTS                                                                                                  GSI DUE TO EXTREME COLD WEATHER 48                            LCO FOR CLASS 1E VITAL.'                                                                                                      GSI            '
INSTRUMENT BUSES IN OPERATING                                                                                                                '
REACTORS l-            49                            INTERLOCKS AND LCOs FOR                                                                                                        GSI I                                            REDUNDANT-CLASS 1E TIE BREAKER 51                            PROPOSED REQUIREMENTS FOR                                                                                                      GSI IMPROVING RELIABILITY OF OPEN CYCLE SERVICE WATER SYSTEMS 57                            EFFECTS OF FIRE PROTECTION                                                                                                    GSI SYSTEM ACTUATION ON SAFETY RELATED EQUIPMENT Amendment F December 15, 1989 l    ?
 
CESSAR nainem. .
TABLE A2-1.(Cont'd)
(sheet 2-of 10)
LIDT OF UNewaOLVED BAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIO ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN
                  . ISSUE                                                          ISSUE gggggg                              ISSUE TITLE                TYPE L
64                    IDENTIFICATION OF PROTECTION            GSI SYSTEM INSTRUMENT SENSING
                                        ~ LINES 66                  STEAM GENERATOR REQUIREMENTS            GSI 75                  GENERIC IMPLICATIONS OF ATWS            GSI EVENTS AT SALEM - ' OPERATIONAL QA PROGRAMS 79                  UNANALYZED REACTOR VESSEL              GSI THEPMAL STRESS- DURING NATURAL CONVECTION COOLDOWN 82                    BEYOND DESIGN BASES ACCIDENTS          GSI IN SPENT FUEL POOL 83                  CONTROL ROOM HABITABILITY              GSI 93                  STEAM BINDING OF AUXILIARY            GSI FEEDWATER PUMPS
"                  .94-                  ADDITIONAL LTOP'FOR LIGHT              GSI WATER REACTOR 99-                  RCS/RHR SUCTION LINE                    GSI INTERLOCKS ON PWRS j-l                  103-                  DESIGN FOR PROBABLE MAXIMUM            GSI PRECIPITATION 105                  INTERFACING SYSTEMS LOCA AT            GSI LWRS 106                  PIPING AND USE OF HIGHLY              GSI COMBUSTIBLE GASES IN VITAL AREAS -- FIRE PROTECTION Amendment F December 15, 1989 hi A_ .                                              1%      -
 
1 C E S S A R RH Wicui.
l 4
l i
j TABLE 12-1 (Conted)                                                                                          l ca.. 2 or 1 3                                                                                        i LIST OF UNRESOLVE:'r SAFETY ISSUES AND RIGE/ MEDIUM PRIORITY GEEERIC ISSUES APPLICABLE TO THE                                                                                        '
RISTIK_f01_.ETAMDARR lRRIGN ISSUE                                                                                                                              ISSUE      '
MUMBER                                  ISSUE TITLE                                                                _
TYPE i
119.1                PIPE RUPTURE REQUIREMENTS                                                                                    GSI/RI      i 119.2                PIPE DAMPING VALUES                                                                                          GSI 119.3                DECOUPLING OBE FROM SSE                                                                      f              GSI/RI        -
119.5                LEAK' DETECTION REQUIREMENTS                                                                                GSI/RI 121          .      HYDROGEN CONTROL FOR LARGE,                                                                                  GSI          ,
DRY PWR CONTAINMENT                                                                                      -
122.2                INITIATING FEED AND BLEED                                                                                    GSI 124'                AUXILIARY FEEDWATER SYSTEM                                                                                  GSI        '
RELIABILITY 3.25.I.3            SPDS AVAILABILITY                                                                                            GSI 125.II.7            REEVALUATE PROVISION TO                                                                                      GSI
,                      AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING LINE BREAK 128                  ELECTRICAL POWER RELIABILITY                                                                                GSI 130                  ESSENTIAL SERVICE WATER PUMP                                                                                GSI FAILURES AT MULTIPLANT SITES 135                  INTEGRATED STEAM GENERATOR                                                                                  GSI ISSUE A-1                  WATER HAMMER                                                                                                USI A-2                  ASYMMETRIC 3 LOWDOWN LOADS ON                                                                                USI RCS A-4                  C-E STEAM GENERATOR TUBE                                                                                    USI INTEGRITY Amendment F December 15, 1989
 
CESSAR M%.n..
L                                            TABLE 12-1 (Cont'd)
(Sheet 4 of 10)
LIST OF UNEESQLYED SAFETY ISSUES AND RIGE/ MEDIUM PRIORITY GENERIC ISSUBS APPLICABLE TO THE SYSTEM S0+ STANDARD DESIGN ISSUB                                                                            ISSUE gggggg                                    IBSUB TITLE                            TYPE A-9                            ANTICIPATED TRANSIENTS WITHOUT            USI SCRAM (ATWS)
A-12                            FRACTURE TOUGHNESS OF STEAM                USI GENERATOR AND RCP SUPPORTS A-13                          SNUBBER OPERABILITY ASSURANCE              GSI A-15                          PRIMARY COOLANT SYSTEM                    GSI DECONTAMINATION'AND STEAM
                                                                                                  /
GENERATOR CHEMICAL CLEANING                              )
A-17                          SYSTEMS INTERACTION                        USI A-24                          QUALIFICATION OF CLASS lE                  USI SAFETY RELATED EQUIPMENT A-25                          NON-SAFETY LOADS ON CLASS lE                GSI POWER SOURCES A-26                            REACTOR VESSEL PRESSURE                    USI TRANSIENT PROTECTION A-29                            PLANT DESIGN FOR REDUCTION OF                GSI VULNERABILITY TO SABOTAGE A-30                            ADEQUACY OF SAFETY REIATED DC                GSI POWER SUPPLIES A-31                            RHR SHUTDOWN REQUIREMENTS                    USI A-35                          ADEQUACY OF OFFSITE POWER                    GSI SYSTEMS A-36                          CONTROL OF HEAVY LOADS NEAR                    USI SPENT FUEL A-40                          SEISMIC DESIGN--SHORT TERM                      USI PROGRAM Amendment F December 15, 1989
 
l)lb!h!h k!I khkr$ ItCATION                                                          <
                                                                                          )
1 TABLE A2-1 (Cont'd)
(Sheet 5 of 10)
LIST OF UNEEBOLVED SAFETY ISSUES AND XIlpE/ MEDIUM                      l PRIORITY GEMERIC ISSUES APPLICABLE TO TEE                    i SYSTEM 80+ STANDARD DESIGM                        ;
ISSUE                        -                                        ISSUE      i NRKSER                                      ISSUE TITLE                TYPE        ;
A-43                        CONTAINMENT EMERGENCY SUMP              USI PERFORMANCE A-44                        STATION BLACKOUT                        USI          -
A-45                        SNUTDOWN DECAY HEAT REMOVAL              USI        >
REQUIREMENTS A-47                        SAFETY IMPLICATIONS OF CONTROL  .      USI          !
SYSTEMS
* A-48                        MYDROGEN CONTROL, MEASURES &            USI EFFECT OF HYDROGEN BURNS A-49                        PRESSURIZED THERMAL SHOCK                USI B-5                        DUCTILITY OF TWO-WAY SLABS &            GSI SHELLS -- STEEL CONTAINMENTS B-36                        DEVELOP DESIGN, TEST, MAINTENANCE        GSI          ,
CRITERIA FOR ATMOSPHERE CLEANUP                        i SYSTEM AIR FILTRATION AND                              l ABSORPTION UNIT.....
B-53                      LOAD BREAK SWITCH                        GSI          $
B-56                      DIESEL GENERATOR RELIABILITY            GSI B-58                      PASSIVE MECHANICAL FAILURES              GSI B-60                      loose PARTS MONITORING SYSTEM-          GSI B-61                      ALLOWABLE ECCS EQUIPMENT                GSI          .
OUTAGE PERIODS                                        !
l-Amendment F December 15, 1989
 
I I
CESSAR Mnine.n..
i 1
TABLE A2-1 (Cont'd)                                                    j (Sheet 6 of 10)
LIST OF UNRESOLYED SAFETY ISSUES AND HIGE/ MEDIUM j
PRIORITY SEMERIC ISSUES APPLICABLE TO THE SYSTEM 80+ STAMDARD DESIGN 4
ISSUE                                                                                      IssvE              :
MUMBER                                    ISSUE TITLE                                    TYPE B-63                        IS01ATION OF IDW PRESSURE                                GSI
!                                  SYSTEMS CONNECTED To THE REACTOR COOLANT PRESSURE BOUNDARY B-66                        CONTROL ROOM INFILTRATION                                GSI MEASUREMENTS
{
C-1                          ASSURANCE OF CONTINU0US LONG                              GSI                    ,
TERM CAPABILITY OF HERMETIC                                                      '
SEALS ON INSTRUMENTATION AND c-                                  ELECTRICAL EQUIPMENT C-2                          STUDY OF CONTAINMENT                                      GSI
'                                  DEPRESSURIZATION BY INADVERTENT SPRAY OPERATION C-4                          STATISTICAL METHOD FOR ECCS                              GSI/RI ANALYSIS C-5                          DECAY MEAT UPDATE                                        GSI/RI
,      C-6                          LOCA HEAT SOURCES                                        GSI/RI
,      C-10                        EFFECTIVE OPERATION OF                                    GSI CONTAINMENT SPRAYS IN A LOCA                                                      +
C-12                        FRIMARY SYSTEM VIBRATION                                  GSI ASSESSMENT                                                                        l l
HF 1.3.4a                    HUMAN FACTORS PROGRAM PLAN -                              GSI                      J 3
MAN MACHINE INTERFACE - LOCAL CONTROL STATIONS
~
j HF 1.3.4b                    HUMAN FACTORS PROGRAM PLAN -                              GSI MAN MACHINE INTERFACE -                                                            i 4
ANNUNCIATORS l
l Amendment F                      i December 15, 1989                I
 
CESSAR inWienio.                                                                                                                                                  ;
l i
i TABLE A2-1 (Cont'd)                                                                                              f (Sheet 7 of 10)                                                                                                l LIST OF UNRESOLVED SAFETY ISSUES AND IIGE/ MEDIUM                                                                                                  }
rmI0mITr aEummle Issues APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGM ISSUE                                                                                                                  -
ISSUE                                .
MUMBER                                            ISSUE TITLE                                                              TYPE                                !
MF 1.3.4c                    HUMAN FACTOR PROGRAM PLAN -                                                                GSI MAN MACHINE INTERFACE -
OPEltATIONAL AIDS HF 1.3.4d                    HUMAN FACTORS PROGRAM PLAN -                                                                GSI MAN MACHINE INTERFACE -                                                                                                          A AUTOMATION AND ARTIFICIAL                                                                                                          ,
INTELLIGENCE i
HF 1.3.4e                    HUMAN FACTORS PROGRAM PLAN -                                                                GSI                                    '
M'AN MACHINE INTERFACE -
COMPUTERS AND COMPUTER
* DISPLAYS l
HF 5.1                        LOCAL CONTROL STATIONS                                                                    GSI HF 5.2                        REVIEW CRITERIA FOR HUMAN                                                                  GSI FACTORS ASPECTS OF ADVANCED                                                                                                        .
CONTROLS AND INSTRUMENTATION HF 8.0                      MAINTENANCE AND SURVEILLANCE                                                                GSI PROGRAM I.C.1                        (1-4) SHORT TERM ACCIDENT                                                                  GSI L                                        ANALYSIS AND PROCEDURES l                                          REVISION I.D.2                        CONTROL ROOM DESIGN REVIEWS --                                                              GSI/TMI PLANT SAFETY PARAMETER DISPLAY CONSOLE I.D.3                        CONTROL ROOM DESIGN -- SAFETY                                                              GSI/TMI SYSTEM STATUS MONITORING I.D.4                        CONTROL ROOM DESIGN STANDARD                                                                GSI l-l                                                                                                                            Amendment F l-                                                                                                                            December 15, 1989 m --
          ,3      9-s=-  p  ,m-a-w-,    y  er m,--,-- ,n-    ,,w  ,,y,a c ,-w.m , ,m,-- , ,- _s  , _ _ _ _ _ _ _ - _                _ - - - - - - - - - - - -------
 
CESSAR !RWienio.
4 4
TABLE A2-1 (Cont'd)
(Sheet 8 of 10)
LIST OF UMRESQLVED SAFETY ISSUES AND RIGE/ MEDIUM PRIORITY GEMERIC ISSUES APPLICABLE TO THE                                                          l SYSTEM 80+ STANDARD DESIGN t
Issue                                                                          ISSUE m                                    TSSUE TITLE                                TYPE t
I.D.5              (1) CONTROL Ro0M DESIGN --                              GSI IMPROVED INSTRUMENTATION RESEARCH ALARMS AND DISPLAY 1.D.5              (2), CONTROL ROOM DESIGN --                            GSI                                [
IMPROVED INSTRUMENTATION
                      'RESEARCH I
I.D.5              (3) CONTROL ROOM DESIdN --                              GSI                                -
ON-LINE REACTOR SURVEILLANCE
  -..                  SYSTEMS -
* I.D.5              (4) CONTROL ROOM DESIGN --                              GSI                                3 IMPROVED INSTRUMENTATION-RESEARCH                                                                                    i I.F.1            QUALITY ASSURANCE - EXPAND                              GSI QUALITY ASSURANCE LIST FOR
* EQUIPMENT IMPORTANT To SAFETY I.F.2              (2) QUALITY ASSURANCE --                                GSI/TMI DEVELOP MORE DETAILED CRITERIA I.F.2              (3) QUALITY ASSURANCE --                                GSI/TMI DEVEIDP MORE DETAILED                                                                        '
CRITERIA I.F.2              (6) QUALITY ASSURANCT --                                GSI DEVII4P MORE DETAILED QA CRITERIA                                                                                    l I.F.2              (9) QUALITY ASSURANCE --                                GSI DEVELOP MORE DETAILED QA CRITERIA Amendment F December 15, 1989
 
c CESSAR !!ntricui.                                                                                                    l I
1 TABLE 12-1 (Cont #d)                                                            ;
l (n ...        or to>                                                      q LIST OF UMRESOLVED S&FETY ISSUES AND RIGE/ MEDIUM PRIORITY GENERIC ISSUES APPLICABLE TO TEF SYSTEM 80+ STANDARD DESIGN ISSUE                                                                                ISSUE ggg333                    _
ISSUE TITLE                              TYPE I                                II.B.1                    $AFETY REVIEW CONSIDERATION --                          GSI/TMI                        .
j                                                          REACTOR COOLANT SYSTEM VENTS II.B.2                    SAFETY REVIEW CONSIDERATION --                          GSI/TMI PIANT SHIELDING TO PROVIDE                                                              !
POST ACCIDENT ACCESS TO VITAL                                                -
AREAS                                                                          ,
SAFETY REVIEW CONSIDERATION --                          GSI/TMI II.B.3 POST ACCIDENT SAMPLING SYSTEM II.C.4                    RELIABILITY ENGINEERING'                                GSI/TMI II.D.1                    COOLANT SYSTEM VALVE --                                GSI/TMI TESTING REQUIREMENT                                                                    ,
II.D.3                    COOLANT SYSTEM VALVES -- VALVE                          GSI/TMI                      ;
POSITION INDICATION II.E.1.1                  AUXILIARY TEEDWATER SYSTEM                              GSI/TMI                      -
EVALUATION II.E.1.2                  AUXILIARY FEEDWATER SYSTEM                              GSI/TMI AUTOMATIC INITIATION AND FLOW INDICATION II.E.4.1                  CONTAINMENT DESIGN --                                  GSI DEDICATED PENETRATIONS II.E.4.2                  CONTAINMENT DESIGN --                                  GSI/TMI                      .
ISOI.ATION DEPENDABILITY
                              'IY.E.4.4                      (1-5) CONTAINMENT DESIGN --                            GSI/TMI PURGING II.F.1                    ADDITIONAL ACCIDENT MONITORING                          GSI/TMI INSTRUMENTATION Amendment F December 15, 1989
 
i t
l l
CESSAR!in hi.                                                                                                                        1 TABLE 12-1 (Cont'd)
(Sheet 10 of 10)
LIST OF UNEESQLTED SAFETY ISSUES AMD RIGE/ MEDIUM PRIORITY GEMERIC ISSUES APPLICABLE TO THE                                                                            ;
SYSTEM 80+ STANDARD DESIGN                                                                              ,
ISSUE                                                                                                                ISSUE            >
33333                                    ISSUE TITLE                                                                TYPE II.F.2                  IDENTIFICATION AND RECOVERY                                                                GSI/TMI FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING II.F.3                  INSTRUMENTATION FOR MONITORING                                                            GSI/TMI            -
ACCIDENT CONDITION                                                                                            >
II.G.1                  POWER SUPPLIES FOR PRESSURIZER                                                            GSI/TMI RELIEF VALVES,. BLOCK VALVES, AND LEVEL INDICATORS i
II.K.1                    (3,4d,6,9,14,15,16,24-28)                                                              GSI      -
i                                MEASURES to MITIGATE SMALL BREAK ICCA'S & 14SS OF FW ACCIDENTS IE BULLETINS II.K.3                    (5,6,8,25,30,31,55) FINAL                                                              GSI/TMI l
RECOMMENDA7 IONS OF B&O TASK FORCE TO MITIGATE ACCIDENTS III.D.3.3                  (1-4) IN-PLANT RADIATION                                                                GSI/TMI MONITORING                                                                                                  1 III.D.3.4                CONTROL ROOM HABITABILITY                                                                GSI l
l l
l                                                                                                                                              1 l
l' Amendment F i
December 15, 1989
 
I CESSAR ini%ario, i
3.0        APPLINRLE NRC USIs/GSIs CROSS-REFERENCED IN NUREG-0933            f Some of the Unresolved and Generic Safety Issue descriptions in              ,
NUREG-0933 include a cross-reference to one or more other issues which address related concerns.        Table A3-1 identifies the            !
cross-references among isFMGJ relevant to the System 80+ Standarc Design.
t k
                                                                            ?
f-l Amendment F A-5                  December 15, 1989      1 j
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i CESSAR RPehr.,,                                                                          ,
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L                                                                                                :
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                                                                                              't 4
                                                                                                  ?
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Amendment F A-6            December 15, 1989                              t-I
                                                  - _ _ _ - - - - - _ _ _ _ _ - _ _ _ _    _  m
 
i CESSAR E bio.
1 i
TABLE A3-1                                .
(Sheet 1 of 2)
MRC USIs/GSIs CR3SS-REFERENCED IN NUREG-0933                l AND APPLICABLE TO TIE SYSTEM 80+ STANDARD DESIGN l-NRC ISSUE                                          CROSS-REFERENCED      !
uvunna                ISSun TItta                    ISSUES 36          LOSS OF SERVICE WATER                  A-45 48          LCO FOR CLASS 1E VITAL                  128              >
INSTRUMENT BUSES IN OPERATING REACTORS                                                  ,
49          INTERLOCKS AND LCO'S FOR                128 REDUNDANT CLASS 1E TIE BREAKER                ,
: 66.          STEAM GENERATOR REQUIREMENTS            A-4              i 79          UNANALYZED REACTOR VESSEL                A.44 THERMAL STRESS-COOLDOWN L
83          CONTROL ROOM KABITABILITY                B-36 122.2        INITIATING FEED AND BLEED                A-45 125.I.3      SPDS AVAILABILITY                        I.D.2 128          ELECTRICAL POWER RELIABILITY            48, 49 130        ESSENTIAL SERVICE WATER PCMP            A-45 FAILURES AT MULTI-PLANT SITES A-30        ADEQUACY OF SAFETY RELATED DC            II.E.1.1 POWER SUPPLIES                                            ,
A-45        SHUTDOWN DECAY MEAT REMOVAL              A-44 REQUIREMENTS I.D.5        (2) CONTROL ROOM DESIGN --              II.F.3 IMPROVED INSTRUMENTATION RESEARCH I.D.5        (4) CONTROL ROOM DESIGN --              II.F.2 IMPROVED INSTRUMENTATION RESEARCH Amendment F pecember 15, 1989
 
I CESSAR Innneuio.
i TABLE R3-1 (Cont'd)                              t (Sheet 2 of 2)
MRC USIs/GSIs CROSS-REFERENCED IN NUREG-0933 AMD APPLICABLE TO TEE SYSTEM 80+ STANDARD DESIGN I      NRC ISSUE                                          CROSS-REFERENCED uDumER                rssuE TITLE      -
IsseEs            l II.C.4      RELIABILITY ENGINEERING                  B-56
    }
4 1
l l                                                        Amendment F December 15, 1989
 
CESSARiinbr.                                                                        l l
j 4.0        TECENICAL ILE80LUTIONS FOR UNRESOLVED AND GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM so+ STANDARD                      j DESIGN                                                                  l 1
This section presents the technical resolution for each safety issue included in this submittal (64 issues).          The resolutions              1 for the safety issues are listed in Table A4-1.          Resolutions for            !
the remaining applicable issues are scheduled for subsequent                      i submittal.
Each issue is structured so as to be independent of other safety issues. However, there are some instances where issues do overlap one another.      Where overlap occurs, as identified in NUREG-0933,    it  is  so  indicated    (see also Section 3.0,                {~
" Cross-references").                                                              ,
As discussed in the Appendix Overview, each issue is composed of four parts:    (1) a ISSUE statement section, which describes the safety concern, (2) a ACCEPTANCE CRITERIA section which discusses                  ,
the applicable NRC guidance and regulations and industry codes,                    ,
standards and/or other relevant requirements,          (3) a RESOLUTION            i section which describes the technical bases for the resolution of
* the  issue    considering  the    System  80+  Standard      Design,  as described within CESSAR-DC or other relevant documentation (e.g.,
s;pecial technical reports) and finally, (4) a REFERENCES section which lists      the  references used      in the  formuli    '.on  of  the      [
Issue Statement, Acceptance Criteria, and kesolution sections of the issue.
t Amendment F A-7                December 15, 1989
 
y o
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      ~' ,-            * ^ = * ** **a
                  ' CES6MR C*RTIFICATION I                                                                                                                                                i t
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                                                                                                                                                  ?
t i
L                                                                                                                                                ;
t i
i THIS PAGE INTENTIONALLY BLANK                                                                          ;
r
                                                                                                                                                  .t 4
6 i
l l-i                                                                                                                                                >
l l-t 1
l I,                                                                                                                                                b Amendment F A-8                  December 15, 1989
 
i I
CESSAR tin %ui..                                                              ,
l 4
TABLE 14-1 (Sheet 1 of 5)
I LIST OF TECEMICAL RESOLUTIONS FOR                        ,
USIs AND GSIs APPLICABLE TO THE SYSTEM 80+
STANDARD DESIGM INCLUDED IM SECTION 4.O I
NRC                                                                          )
ISSUE MUMBER                            ISSUE TITLE                    PAGE No. l 3                    SETPOINT DRIFT IN INSTRUMENTATION          A-9        l 22                  INADVERTENT BORON DILUTION EVENTS          A-12 23                  REACTOR COOLANT PUMP SEAL FAILURES          A-14        -
29                  BOLTING DEGRADATION OR FAILURES IN          A-17 NUCLEAR PLANTS 36                  LOSS OF SERVICE WATER                      A-19 45                    INOPERABILITY OF INSTRUMENTS              A-20 DUE TO EXTREME COLD WEATHER r
48                    LCO FOR CLASS 1E VITAL INSTRUMENT          A-24 BUSES IN OPERATING REACTORS                            ,
49                    INTERLOCKS AND LCOs FOR CLASS 1E          A-26 TIE BREAKERS                                            ;
i              51                    PROP 03ED REQUIREMENTS FOR IMPROVING      A-28 RELIABILIT1 0F OPEN CYCLE SERVICE WATER SYSTD3 1
57                    EFFECTS OF FIRE PROTECTION SYSTEM          A-31        -
ACTUATION ON SAFETY-RELATED EQUIPMENT l              64                  IDENTIFICATION OF PROTECTION              A-33
: j.                                  SYSTEM INSTRUMENT SENSING LINES 66                  STEAM GENERATOR REQUIREMENTS              A-35 79                    UNANALYZED REACTOR VESSEL THERMAL          A-37        f STRESS-COOLDOWN 82                    BEYOND DESIGN BASES ACCIDENTS IN          A-39 SPENT FUEL POOLS Amendment F December 1?,, 1989 l
l
 
CESSAR iminem .
                            - TABLE A4-1 (Cont'd)                                    .
(Sheet 2 of 5)                                        )
LIST OP TECINICAL RESOLDFIONS POR i
USIs hMD GSIs EPPLICABLE TO THE SYSTEM 80+
STAMD&Ep DESIGM_IMCLUgED IM SECTION 4.O NRC                                                                            . l ISSUE MUMERE                          ISSUI TITLE                  PAGE MC.
83                CONTROL ROOM HABITABILITY                  A-41 t
93                STEAM BINDING OF AUXILIARY                A-43                  '
FEEDWATER PUMPS 103              DESIGN FOR PROBABLE MAXIMUM                A-46 PRECIPITATION 106              PIPING AND USE OF HIGHLY COMBUSTIBLE      A-48 GASES IN VITAL AREAS -- FIRE
* PROTECTION 119.1            PIPE RUPTURE REQUIREMENTS                  A-50 119.2            PIPE DAMPING VALUES                        A-52 119.S            DECOUPLING OBE FROM SSE                    A-55 122.2            INITIATING FEED AND BLEED                  A-58 124              AUXILIARY TEEDWATER SYSTEM                A-60      %        i RELIABILITY 125.I.3          SPDS AVAILABILITY                          A-62 128              ELECTRICAL POWER RELIABILITY              A-64 130              ESSENTIAL SERVICE WATER PUMP              A-65 FAILURES AT MULTI-PLANT SITES A-2              ASYMMETRIC BLOWDOWN LOADS ON RCS          A-67 A-9              ANTICIPATED TRANSIENTS WITHOUT            A-69 SCRAM (ATWS)
A-12            FRACTURE TOUGHNESS OF STEAM                A-72 GENERATOR AND RCP SUPPORTS Amendment F December 15, 1989 I
                                              .-.      -_              --      y
 
C E S S A R !H ane m ..
TABLE 14-1 (Cont'd)
(Sheet 3 of 5)
LIaT or enCMmICAL masoLpTIoms rom DRIs AMD BSIs 1PPLICABLE TO TIE SYSTEM 80+
ST&MD&RD DESIGN IMCLUDED IM SECTION 4.0 unc
* ISSUB ggMBEg                        Issen TITLa                23aa uo.
A-13              SNUBBER OPERABILITY ASSURANCE            A-74 A-25              NON-SAFETY LOADS ON CLASS 1E            A-77 POWER SOURCES A-26              REACTOR VESSEL PRESSURE TRANSIENT        A-80 PROTECTION A-29              PLANT DESIGN FOR REDUCTION OF          A-84 VULNERABILITY To SABOTAGE A-30              ADEQUACY OF SAFETY RELATED DC          A-88 POWER SUPPLIES A-31              RHR SHUTDOWN REQUIREMENTS              A-90 A-36              CONTROL OF HEAVY I4 ADS NEAR            A-94 SPENT FUEL A-43              CONTAINMENT EMERGENCY SUMP              A-97 PERFORMANCE A-45              SHUTDOWN DECAY HEAT REMOVAL              A-101 REQUIREMENTS A-41              PRESSURIZED THERMAL SHOCK                A-103        >
B-60              LOOSE PARTS MONITORING SYSTEM          A-106 C - i.            STATISTICAL METHODS FOR ECCS            A-109 ANALYSIS C-5              DECAY HEAT UPDATE                      A-111 C-12              PRIMARY SYSTEM VIBRATION ASSESSMENT    A-113 Amendment F December 15, 1989 l
 
c CESSAR !!nt!,en..                                                                                                          l i
I TABLE 14-1 (Cont'd) i (Sheet 4 of 5)
LIST OF TECENICAL RESOLUTIONS FCE                                                      1 l
USIs AMD GSIs APPLICABLE TO THE SYSTEM 80+
SPAMDERD DESIGN 12CLUDED IM SECTION 4.0 NRC ISSUE                                                                                                                      '
NUMBES                                              ISS'JE TITLE                          PAGE NO.
HF 1.3.4a                            HUMAN FACTORS PROGRAM PIAN -                          A-115                          i I4 CAL CONTROL STATIONS HF 1.3.4b                            HUMAN FACTORS PROGRAM PLAN -                          A-116 ANNUNCIATOR SYSTEMS HF 1.3.4c                            HUMAN FACTORS PROGRAM PLAN -                          A-117 OPERATIONAL AIDS HF 1.3.4d                            HUMAN FACTORS PROGRAM PLAN -                          A-117 AUTOMATION AND/OR ARTIFICIAL            +
INTELLIGENCE SYSTEMS HF 1.3.4e                            HUMAN FACTORS PROGRAM PLAN -                          A-117 COMPUTERS AND COMPUTER DISPLAY TECHNOLOGY                                                                        ,
HF 5.1                                IDCAL CONTROL STATIONS                              A-120 HF 5.2                                REVIEW OF CRITERIA FOR HUMAN                        A-121 FACTORS ASPECTS OF ADVANCED INSTRUMENTATION AND CONTROLS (ANNUNCIATORS)
I.C.1                                SHORT TERM ACCIDENT ANALYSIS                        A-122 AND PROCEDURES REVISION                                                            ,
I.D.5                                  (2) CONTROL ROOM DESIGN --                          A-124 IMPROVED INSTRUMENTATION RESEARCH - PLANT STATUS AND POST-ACCIDENT MONITORING l
I.D.5                                  (3) CONTROL ROOM DESIGN --                          A-127
('
l ON-LINE REACTOR SURVEILLANCE l                                          SYSTEMS i
Amendment F December 15, 1989 l
L r      _ _ _ _ _ _ _ _ _ _ - _ _ .                    .-. . _ .        . - .  -_  . . - .-        - --..__ - . - --
 
i CESSAR Mnen..
I 1
1 i
TABLE 14-1 (Cont *d)                                                                                                .
J (Sheet 5 of 5) l LIST OF TECINICAL RESOLUTIONS FOR USIs AND GSIs APPLICABLE TO THE SYSTEM 80+                                                                                          ;
STAMDARD DESIGN IMcLUDED IN 82C'110N 4.0 l
      **C                                                                                                                                              i Issen 333333                          ISBUE TITLE                                                                          PAGE No.
l    I.D.5                (4) CONTROL ROOM DESIGN --                                                                      A-130 l                          PROCESS MONITORING
* INSTRUMENTATION II.B.1              SAFETY REVIEW CONSIDERATION --                                                                  A-133        ,s REACTOR COOLANT SYSTEM VENTS A-135                      i II.B.3              SAFETY REVIEW CONSIDER 6 TION --
POST ACCIDENT SAMPLING SYSTEM II.C.4              RELIABILITY ENGINEERING                                                                          A-138 II.D.1              COOLANT SYSTEM VALVES --                                                                        A-141                      ,
PERFORMANCE TESTING REQUIREMENTS l                                                                                                                                                      ,
A-143 II.D.3              COOLANT SYSTEM VALVES -- VALVE POSITION INDICATION II.E.1.1            AUXILIARY FEEDWATER SYSTEM                                                                      A-146 EVALUATION II.E.1.2            AUXILIARY FEEDWATER SYSTEM                                                                      A-149 AUTOMATIC INITIATION AND FLOW INDICATION II.F.2              INSTRUMENTATION FOR DETECTION                                                                    A-151 0F INADEQUATE CORE COOLING II.F.3              INSTRUMENTATION FOR MONITORING                                                                  A-154 ACCIDENT CONDITIONS II.G.1                POWER SUPPLIES FOR PRESSURIZER                                                                  A-157 RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATORS l
l Alnendment F December 15, 1989
 
l I
CESSAR WMeann l
003t                SETPOINT DRIFT IN INETRUMENTATION                                                                      ,
i i
ISSUE                                                                                                                                    ,
Generic Safety Issue (GSI) 003 in NUREG-0933 (Reference 1),                                                                              i addresses drift in safety-related instrumentation and controls setpoints and the potential for a delay in initiation of a                                                                              i safety-related system or component.                                                                                                      .
Setpoint drift can be defined as a change in the input-output relationship of an instrument over a period of time. Setpoint drift can occur as a result of a number of factors including component  failure,                            instrumentation error and environmental conditions.      Setpoint                                drift    primarily    affects                          analog instrumentation rather than digital instrumentation (which is less sensitive to the environmental effects of temperature,                                                                              '
humidity, etc.).                        Safety-related instrumentation and controls systems use setpoints as a means of determining when to initiate a safety function. Should an unplanned change in the setpoint of l-  a safety-related component occur (i.e., setpoint drift) the actual value of the measured parameter at whi'ch a particular                                                                            >
  . action is specified to occur will be altered. This phenomenon can result in the delay in the initiation of a safety function.
A number of Licensee Event Reports (LER's) were reviewed by the
* NRC    whivh dealt with setpoint drift                                        in      safety-related instrumenkition and controls. Subsequsntly, many of these LER's were determined to have reported setpoint drift in safety-related instruments bsyond their permissible technical specification
!    limits. Therefore, the NRC determined that it was necessary to l
provide industry with additional guidance, Regulatory Guide 1.105 l    (Reference 2), which could be utilized in establishing and maintaining safety-related instrument setpoints. In conjunction with the NRC work, industry developed a standard, ISA $67.04-1987 (Reference 3) for safety-related instrument setpoints. This revised standard replaces ISA S67.04-1982 endorsed by Regulatory Guide 1.105, Revision 2.
SCCEPTANCE CRITERIA The acceptance criteria for the resciution of GSI 003, are that safety-related          instrumentation and controls systems                                                                  and I    components which use setpoints as a means of initiating their safety functions shall: (1) establish and maintain the setpoints using the. guidance in Regulatory Guide 1.105, Revision 2, (with the exception of ISA S67.04-1982) and (2) conform to the criteria identified in ISA S67.04-1987.
l l
!                                                                                        Amendment F A-9                  December 15, 19.89
 
CESSAR E!n*nem.
Specifically, a setpoint shall be established such that its selection shall allow sufficient margin between the setpoint and the technical specification limit to account for the expected instrument drift. In particular, the setpoint selection shall consider the expected environmental conditions.
BEBOLUTION The    System    80+  Standard  Design    includes    safety-related instrumentation and controls which use setpoints to begin safety functions (See CESSAR-DC, Section 7.0) . Setpoints for the safety related systems and components      (e.g., Plant Protection System (PPS)) are established and maintained in accordance with the guidance given in Regulatory Guide 1.105 and conform to the criteria identified in ISA-S67.04-1987 (See CESSAR-DC, Section 7.1.2.27).
The environment considered when determining errors is the most detrimental realistic environment calculated or postulated to exist up to the longest time of the required Reactor Trip or Engineered Safety Feature Actuation. This environment may be different for diffeRent events analyzed.          For the    setpoint  ,
calculation, the accident      environment  error  calculation  for  i process equi? ment uses the environmental conditions up to the longest requ:. red time of trip or actuation that results in the largest errors, thus providing additional conservatism to the resulting setpoints. For additional detail on safety-related              .
instrumentation and control setpoints see CESEAR-DC,          Section  l 7.1.2.27.  (Note actual equipment setpoints are determined during      '
plant construction, when specific equipment is purchased, and are based on the safety analysis setpoints and setpoint methodology described in CESSAR-DC.)
Although setpoint drift is expected to be negligible because of digital protective systeins, periodic surveillance tests designed to detect input parameter drift and            setpoint changes are performed. In addition, automatic testing of the PPS bistable trip function is provided and PPS trip setpoints are monitored by        1 the . Data Processing System as described in CESSAR-DC, Sections 7.2.1.1.9 and 7.7.1.8.2.I.
Setpoints which are used to initiate these plant safety functions are established and maintained by implementing the requirements of ISA-S67.04-1987 and therefore meet the intent of Regulatory          '
Guide 1.105, Rev 2. Since the guidance and requirements are met, this issue is resolved for the System 80+ Standard Design.
Amendment F        I A-10                  December 15, 1989  {
l
 
<  CESSAR !!nificui.
REFERENCE 5
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. Regulatory Guide 1.105, Revision 2, " Instrument Setpoints For  Safety-Related. Systems",    U.S. Nuclear Regulatory Commission, February 1986.
  -3. Standard    ISA-567.04-1987,      "Setpoints    for    Nuclear Safety-Related    Instrumentation Used In      Nuclear    Power Plants", Instrument Society of America.
4 i
s t
                                                                                ?
rM Amendment F            ;
A-11                December 15, 1989
 
i CESSAR !!nincano.
1 022:  INADVERTENT BORON DILUTION EVENTS ISSUE                                                                                                J Generic Safety Issue        (GSI)  022 in NUREG-0933 (Reference 1),
addresses    the    possibility    of      core criticality during cold shutdown conditions because of an inadvertent boren dilution event.
1 Inadvertent boren dilution events have occurred at PWR's during                                      I I
maintenance and refueling periods. If the boron in the RCS is sufficiently diluted and the reactor core is near the beginning of life, there is the potential for core criticality with all rods inserted (i.e., during cold shutdown conditions).
l l    The NRC and others performed a variety of studies of the consequences of an inadvertent boron dilution event. The conclusions of the NRC assessment along with other studies were                                      1 (1) that the consequences of an unmitigated boron dilution event,                                    )
although undesirable, are not severe enough to warrant backfit of                                    '
l additional    protective    features at operating              plants      and  (2)
Standard Review Plan        (SRP)    Section 15.4.6          (Reference 2)        is adequate for plants presently undergoing license review.
ACCEPTANCE CRITERIA The acceptance criterien for GSI 022 is that new plants shall minimize the consequences of inadvertent boron dilution svents by                                    i meeting the intent of SRP Section 15.4.6.                    Specifically, when                      ;
performing a safety analysis to evaluate the consequences of an inadvertent boron dilution, plant designers should consider: (1) i    design limits for maximum RCS pressure and minimum DNBR, (2) moderate frequency events in conjunction with a single failure or operator error and their possible effects on fuel integrity and radiological dose calculations, (3) and time limits specified for each mode of plant operation, if operator action is required to terminate an inadvertent boron dilution.
RESOLUTION i
i    As part of the design process for the System 80+ Standard Design, Safety Analyses are performed.            These analyses address a variety of design bases events including inadvertent boron dilution (see
. CESSAR-DC, Section 15.4.6). Furthermore, these analyses consider SRP Section 15.4.6 criteria including, design limits (e.g.,
maximum RCS pressure and minimum DNBR),                  a single failure in conjunction with moderate frequency events, and the impact of a single    failure    or  operator        error  en      fuel  integrity      and radiological dose calculations.
Amendment F A-12                        December 15, 1989
 
1 l
CESSAR MNano.
The safety analysis also considers the time limits required for an operator to terminate an inadvertent boron dilution for a particular plant mode of operation identified in SRP, Section                    i 15.4.6. For example, indication of a boren dilution ovent in the                !
cold shutdown mode of operation is provided to the operator by                  j the boron dilution alarm logic within the Nuplex 80+ Advanced As described in CESSAR-DC, Section I
control complex (ACC).
7.7.1.1.10, the boron dilution alarm logic would detect an                        ,
inadvertent boron dilution event by monitoring the neutron flux                  j indications provided by the ex-core detector instrumentation while in plant modes 3 through 6. Alarm signals are generated by the non-safety-related discrete indication and alarm system                    ,
(DIAS) and the data processing system (DPS) alerting the operator. Reliance on the boron dilution alarm logic is credited in the CESSAR-DC Chapter 15 Safety Analysis as the annunciator of              i the event and assures that the 15 minute and 30 minute criteria for loss of shutdown margin are met.
The System 80+ Standard Design includes a safety analysis which              I c demonstrates that the consequences of an inadvertent boron
* i dilution during cold shutdown are minimized. Furthermore, clear and concise indicaticn and alarm instrumentation is provided to an operator via the Nuplex 80+ ACC which is considered in the                  ,
safety analysis. Therefore, the intent of SRP Section 15.4.6 is met and this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Prioritization of Generic Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0800, " Standard Review Plan for the Review of Safety
                                                        "      LWR Edition, Analysis Reports for Nuclear Power Plants        ,
U.S. Nuclear Regulatory Commission.
: 3. Letter to R. T. Curtis (NRC) from N. S. DeMuth ( LANL) ,
        " Analyses  of    Unmitigated    Boron    Dilution        Events",
November 18, 1981.
Amendment F A-13                    December 15, 1989
 
CESSAR !!ninew.,                                                                    ,
i 023:          REACTOR COOLANT PUMP SEAL FAILURES                  .
5 IS8HE Ceneric Safety Issue (GSI) 023 in NUREG-0933 (Reference 1),                        l addresses the high rate of Reactor Coolant Pump (RCP) seal failures that challenge the makeup capacity of the ECCS in PWRs which could result in a small-break loss-of-coolant-accident (IOCAs) and possibly result in core damage.
WASH-1400,    (Reference 2) indicated that breaks in the reactor coolant pressure boundary having an equivalent diameter in the range of 0.5 to 2 inches were a significant cause of core melt.
Since then, a study has shown that comparable break flow rates have resulted from RCP seal failures at a frequency about an order of magnitude greater than the pipe break frequency used in WASH-1400.      Thus, the overall probability of core melt due to small-size breaks could be dominated by events such as RCP seal failures.
While proper design and testing of the RCP shaft seal                        is important, the most significant area of concern is the capability of the seals to withstand a combined loss of component cooling water (CCW) and saal injection (SI), which could occur during station blackout (SBO) conditions.                                                  ,
!'  ACCEPTANCE CRITERIA l
The. acceptance criteria for the resolution of GSI 023 are that                    ,
i the plant designer should provide for greater RCP shaft seal l    integrity than has been exhibited by the seal design on some l    operating plants.                Seal integrity can be obtained by using improved design features, such as, better pump and shaft seal design, and/or improved seal auxiliary support systems (component cooling water and seal injection systems).
The objective of improved RCP shaft seal integrity is to limit the possibility of a small-break LOCA (which might lead to core damage) resulting from a RCP shaft seal failure.                In particular, the susceptibility of the auxiliary systems to failure because of a station blackout (SBO) should be addressed.
RESOLUTION The System 80+ Standard Design minimizes the possibility of core damage resulting from a small-break LOCA event caused by a RCP shaft seal failure by assuring seal integrity.                  Seal integrity is ensured by seal and support systems design which address, as described below, susceptibility to station blackout.
Amendment F A-14              December 15, 1989
 
CESSAR Emincara The Reactor Coolant Pump shaft seal design for System 80+ is a CE-KSB design. This design employs 3 cartridge type, hydrodynamic seals, including 2 equally staged seals and a third stage used as a vapor seal (see CESSAR-DC, Section 5.4.1.2). With this type of                                    ,
seal arrangement, all three seals must fail before leakage from the reactor coolant system will occur. In addition to the seal design, the RCPs and integral shaft seals are factory performance                                  >
tested. Factory testing for a variety of operating conditions demonstrates that the RCP shaft seal design is satisfactory.
(Examples of factory test and actual operating experience will be discussed    later.)      The RCP shaft seals are                  operated              in conjunction with two support systems, the RCP Seal Injection (SI) and component Cooling Water (CCW) Systems.
The RCP shaft seal and supporting systems provide improved RCP seal integrity.        For example, to maintain real injection under SB0 conditions, the System 80+ Standard design incorporates an l  on-site alternate AC (AAC) power source.                      As described in CESSAR-DC Section 8.1.4.2, the installation and design of the AAC                                ,
{:  source is in compliance with the intent of Regulatory Guide 1.155 (Reference 3).      The AAC would be used to power the charging pumps
  ! which supply seal injection water to cool the RCP shaft seals.
In the unlikely event that AAC is not availabe and a SB0 occurs the shaft seals are capable of limiting leakage to a maximum of 8                                  '
gpm per pump without cooling.                                                                      :
The 8 gpm per pump is based upon operating and test experience with  hydrodynamic      shaft    seals        used in CE designed plants.
Earlier CE plants used seals designed by Byron Jackson (B-J) while the System 80 (Palo Verde) units contain CE-KSB seals. (The System 80+ reactor coolant pump shaft seal design is the same as for the System 80 design, i.e., CE-KSB seals.)                  Both the B-J and CE-KSB seal designs are hydrodynamic which operate in the same manner, contain similar materials and behave similarly under abnormal conditions.          The capability of the B-J seals to
,  withstand SBO conditions has been demonstrated by testing two l  production    seal    cartridge          assemblies. In both      tests,          seal leakage was less than 2 gpm for an 8 hour SBO test duration.
The capability of the CE-KSB seals has been further demonstrated by the following operating event at the Palo Verde plant.                                  In April 1986, Palo Verde Unit 2 RCP 2B experienced a loss of CCW and SI for three hours.          During this three hour interruption the pump was operated for 10 minutes before it was tripped.
This resulted in the pump seals being exposed to RCS hot standby temperature conditions.          No loss of seal function occurred and there was no measurable increase in leakage to the containment.
Amendment F A-15                December 15, 1989
 
k
_    CESSAR W Ln..
i r          Following this event, the affected RCP was placed back ir.to service without inspection of the seals. The RCP was operated
;          for several months prior to a normal refueling and maintenance l-          whutdown.
Although this event time duration was shorter than the 8-hour coping criteria for SBO, the event was potentially more severe than a SBO because the pump was operated without cooling for 10                                          ,
minutes prior to shutdown.              This type of operation rapidly subjects the seals to RCS temperatures as hot coolant flows upward through the seals.            Also, there is no mechanism for l
rejecting the heat built up in the seals due to rubbing friction.
;          In addition, to the Palo Verde event, seal integrity was further
!          demonstrated by a 500 hour shop RCP endurance test during which various loss of seal cooling events were tested.
The System 80+ RCP seal design is based on techno:ogy proven in operating plants. In addition, the SI and CCW support Systems for                                        i the System 80+ Standard Design are powered by an additional AAC power source.        This alternate AC source, under SBO conditions, f
,          sustains electrical power to the charging pumps and component L. .        cooling water pumps which enables the RCP seal cooling and seal 1          leakage to be maintained within acceptable limits. Because RCP                                            i shaft seal cooling and shaft seal leakage are maintained within u .'.:
acceptable limits,          excessive seal leakage will not occur.
$          Therefore, reactor core damage will not result from a loss of RCS                                        -
water inventory due to excessive seal leakage and thus, this
#            issue is resolved for the System 80+ Standard Design.
PEFERENCES
: 1.        NUREG-0933, "A Status Report On Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.        WASH-1400  (NUREG-75/014),    " Reactor      Safety          Study,                  An    ;
Assessment of Accident Risks in Commercial Nuclear Power Plants", U.S. Nuclear Regulatory Commission, October 1975.
1
: 3.        Regulatory Guide 1.155, " Station Blackout",                U.S.                Nuclear Regulatory Commission, June 1988.
l l
I l
l Amtmdment F A-16                          December 15, 1989                      ,
l I
 
1 CESSAR Enlificui.=
029;    BOLTING DEGRADATION OR FAILURE IN NUCLEAR POWER PLANTS                                                            ;
1 u
ISSUE Generic Safety Issue -(GSI) 029 in NUREG-0933 (Reference 1) ,
addresses bolting degradation within safety-related components and support structures and its impact on the integrity of the                                                                    ,
reactor coolant pressure boundary.                                                                                              j The most crucial bolting applications are those constituting an integral part of the primary pressure boundary such as closure                                                                    i studs and bolts on reactor vessels, reactor coolant pumps, and                                                                  !
steam generators.              Degradation of these bolts or studs could                                                        ]
result    in      the      loss    of      reactor      coolant.                                      Other    bolting      ,
applications such as component support and embedmont anchor bolts or studs are essential for withstanding transient loads created during abnormal or accident conditions.
Review      of        operating        experience          demonstrated                                      that    the    -
owner-operator's            maintenance practices significantly affect bolting degradation.
ACCEPTANCE CRITERIA                                                                                              -
The acceptance criteria for the resolution of GSI 029 are that proven bolting designs, materials, and fabrication techniques shall be employed.                  In addition,            reactor coolant pressure                                            :
boundary (RCPB) bolting shall meet the requirements of ASME Code, Section    III        (Reference 2). Also,                  for                                    RCPB  bolting      the owner-operater shall use established                                industry                                practice      in l  developing maintenance, assembly, and disassembly procedures.
!  Furthermore for RCPB and its support bolting, inservice j
inspection shall meet the requirements of ASME, Section XI (Reference 2).
RISCLUTION Bolting degradation of RCPB bolts is primarily an operating plant issue since most of the degraded bolts have resulted from poor maintenance practices.                Bolting integrity is assured by the designer through the initial specification of proven bolting materials        and        installation              requirements,                                      and      by    the    .
,  owner-operator through the use of acceptable maintenance and inspection practices.                                                                                                            .
I ic Amendment F A-17                                                    December 15, 1989 b                                                            . __ _ - _ _ _ _ - _ _ __ _ _ _ _ __ _ _
 
          - CESSAR !!Me m.                                                                                                I l
4 f
For-the System'80+ Standard - Design RCPB, only proven materials -
                ~
.__        for the-specific-application and' environment are employed, having                                              i l'        been selected: after evaluation of the potential for - corrosion wastege      and    intergranular      stress: corrosion                      cracking        (see-            l
          . CESSAR-DC, . Section 5.2.3.2.1).            Also, the RCPB components and their- integral bolts,            including the                        unctor vessel, steam generators,        reactor coolant pumps ano t ipkg are fabricated, tested, and installed in accordance witA AtM Code,. Sections III                                                l and XI.          Finally, the owner-operator must perform periodic u          inservice inspection in accordance with ASME Code Section XI (see                                              1
          - CESSAR-DC, Section, 5.2.1.2, Table 5.2-1, and Section 5.2.4).                                    In            .
          - addition, for crit
* Mal pressurs boundary _ applications such as the                                          l
          - reactor-vessel. head closure, redundant seals and leak monitoring further assure the integrity of the RCPB.
For major' component support bolting applications (e.g., reactor vessel, steam generator, etc.), the bolts are -designed and fabricated to ASME Section III requirements with proven materials                                          '
chosen 'for the- specific application and environment                                          (see CESSAR-DC, Section 3.9.3)
In summary, the bolting for RCPB components and supports are selected consi.dering their particular application and are fabricated, tsaced and        installed in accordance with ASME Sections III.      Furthermore, the    owner-operator-is required to comply with I.        ASME      Code,. Section      XI    for the            performance of                inservice inspection. Therefore,        this issue is resolved for the System 80+                                      i Standard Design.                                                                                              ;
REFERENCES
: 1.        NUREG-0933, "A Status Report on Unresolved Safety Issues,"
U.S. Nuclear Regulatory Commission, April 1989.
: 2.        American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III (Nuclear) and Section XI, American
!                    Society of Mechanical Engineers.
i; i
                                                                                                                          )
i i
Amendment F A-18                                  Decemoer 15, 1989
        -                        - - -      ..      -    _. . - - . ~ - - -                    . . -    .  .-
 
.2 CESSAR!ah ma 036:  LOS8 OF SERVICE WATER ISSUE Generic Safety Issue        (GSI)  036 in NUREG-0933              (Reference 1) ,
identifies the potential for the loss of both redundant trains of service water caused by the failure of a - non-safety - system or, component.
Calvert Cliffs Unit 1 experienced a loss of both redundant trains of service water . when the- Station Service Water System (SSWS) became. air. bound        as  a    result            of  the  failure    of  a non-safety-related. instrument air compressor aftercooler.                      The significance of this event lies in the fact that it involved two fundamental aspects in tne design of safety-related systems: (1)
      . interaction between safety and non-safety-related systems and                      _
components, and-(2) single. failure of redundant safety systems.                    n ACCEPT 1McB CRITERIA The acceptance criterion for the resolution of GSI 036 is that the design of the SSWS shall be such that: (1) the potential for the loss of its safety function through system interaction with a non-safety-related system be minimized, and (2) the potential for                      ,
a single failure of the SSWS shall be minimized.                    Specifically, the SSWS .& hall be designed to meet the intent of SRP Sections
      .9.2.1, Rev. 4, and 9.2.2, Rev. 3 (Reference 2).
RESOLUTION Aspect number (1) of this issue is resolved by the System 80+
Standard Design because its safety-related SSWS only cools the component cooling water heat exchangers which are                              also safety-related.      Hence    there    are          no  safety-to-non-safety interfaces (see CESSAR-DC, Section 9.*;.1).                A.spect number (2) is is.resolvad by the System 80+ Standarc Design because its SSWS is comprised of two physically separate, independent full capacity divisions. Thus,  a single failure does not                  impair system effectiveness (see CESSAR-DC, Section 9.2.1. 3) .                Since the SSWS meets the existing requirements and NRC guidance, this issue is
: 4.      resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
  )          U.S. Nuclear Regulatory Commission, April'1989.
: 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Pcwer Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.                                              :
Amendment F A-19                          December 15, 1989 b                                                          _
 
3 CESSAR1!nLm                                                                                                                                  l l
;'                                                                                                                                                l e
045t      INOPERABILITY OF INSTRUMENTATION DUE TO EXTD MR COLD WEATHER ISSUE Generic Safety Issue                              (GSI)    045 in NUREG-0933                          (Reference          1),              l addresses the potential for safety-related equipment instrument lines to become inoperable as a result of freezing or reaching the . precipitation                    (i.e.,          solidification) point of the sensing fluids.
1 Typical safety-related systems employ pressure and level sensors which use small bore                                  instrumentation lines.                          Most          operating i
plants ,contain safety-related equipment' and systems, parts of l
which are exposed to the ambient environment.                                                                    These lines generally contain                        liquid            (e.g.,                      borated  water) which is susceptible to freezing.                                Where  systems                    or components          and their        ."
associated                    instrumentation                are                      exposed    to            sub-freezing temperatures, heat tracing and for insulation is used to minimize l
the effects of cold temperatures,                                                                                                          ,
These sensing. and instrumentation lines are of concern because, should. they freeze, they may prevent a safety,-related system or.                                                                      1 4    component-from performing its safety function. For example, an incident occurred at a plant _ wherein the heat tracing system                                                                            ;
surrounding sensing                          lines          and level transmitters for the                                            ' l Refueling Water Storage Tank (RWST) failed during sub-freezing                                                                          j weather.' The failure of the heat tracing system resulted in freezing of the sensing lines and associated level transmitters causing a loss - of all four RWST instrumentation channels, which could have resulted in tho failure of the Emergency Core Cooling System, thus jeopardizing plant safety.
Because of,the possibility                                of a safety-related system failure, the NRC issued additional guidance given in Regulatory Guide 1.151 (Reference 2), to supplement the existing guidance and requirements which include the Standard Review Plan (SRP) Section 7 .1, - 10 CFR 50,                    Appendix A and industry Standard ISA-67.02, (References                3,      4 & 5, respectively).                                    Regulatory Guide 1.151 specifically                      addresses              the    prevention                      of        safety-related                i instrument sensing line freezing and includes design issues such as diversity, independence, monitoring and alarms.
                                                                                                                                              ^
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 045 is that the fluid in safety-related equipment instrument sensing line.s shall be protected from freezing and maintained above the precipitation point.
Amendment F A-20                                    December 15, 1989 z    _ . . _ . _ _ _ _ _ _ _ _ _- .          _._____ __            _ __ _ _ _ _ _ _ _ _ _                _ . _ _ . _  _ __    ._ . __
 
i 1
.      CESSARiinineim, The protection of safety-related equipment instrument sensing                        ]
lines  from    freezing -can      be    accomplished    by  providing environmental control systems which moet the requirements of 10 I
CFR 50, Appendix-A (GDCs) and industry standard ISA-S67.02, and the intent of Regulatory Guide 1.151, and SRP Sections 7.1, (Rev. 3), 7.1 Appendix A,        (Rev. 1), 7.5, (Rev. 3), and 7.7,                  l (Rev. 3).
Also,  should  environmental control prove to be limited,                        5 alternative forms of sensing line protection such as heat tracing                  ,
and/or insulation can be used.- (The use of heat tracing and/ or                    ,
insulation is not anticipated for the System 80+ Standard Design, however it is' an acceptable alternative to environmental l      control.)
RESOLUTION
~
The System . 80+ Standard Design incorporates instrument sensing              y*
lines    in  safety-related      systems      and  components. All safety-related systems and components used in the System 80+
Standard Design, including instrument sensing lines, are located in a-temperature controlled environment which is maintained above-the freezing (or precipitation) point of the contained fluid.
Because each building has a particular set. of environmental control requirements,- (e.g., slight negative pressure for the                7 fuel building- to      assure no airborne- radioactive material                >
leakage), it has its own ventilation system as shown in Table 1.C                  ,
below. Thus, different General Design Criteria (GDCs) apply to                ."'
various building ventilation designs. In addition to meeting the particular GDC, the guidance given in Regulatory Guide 1.151,                    ;
with respect to redundancy, diversity and monitoring and alarms has been considered.      Discussions of the specific building ventilating systems, their design bases, -including the specific GDCs they address, are provided in CESSAR-DC, Section 9.4.
Limiting conditions for operation for the building ventilation systems which provida- environmental control for the buildings which house these safety-rele.ted systems require that the plant be placed'in a safe shutdown condition should the temperatures in these buildings exceed specified ranges. This assures that the safety-related systems and components are not exposed to freezing or adverse conditions (see CESSAR-DC, Section 16.0). This is consistent with the guidance given in Regulatory Guide 1.151, the applicable SRP Sections,    and the recommendations of industry                  i codes and standards, including ISA-S67.02.
The System 80+ Standard Design uses an IRWST which is inside the containment building wherein the air temperature is maintained between 60 and 90 degrees F (see CESSAR-DC, Section            9.4.5.1).
Furthermore,    Safety-related      equipment    is located    in  the containment building and other areas, wherein the air temperature is maintained above freezing or the precipitation point for sensing line fluids as identified in Table 1.0.
Amendment F A-21                  December 15, 1989
 
x  .
CESSAR Miriemor TABLE 1.0 n
Anticipated Temperature          CESSAR-DC Area                    annee (Decrees (F))              Baction Control Building                73 to 78 (control room)          9.4.1.1 Fuel Building                    60 to 80                        9.4.2 Auxiliary Building,                60 to 100                        9.4.3.1 Rad.: Waste
          . Building                          50 to  100                      9.4.3.1 i            Diesel Building =                60  to  110                      9.4.4.1 a            Containment Building-              60  to  120 (normal oper.)      9.4.6.1 Turbine Building                  50 to  110                      9.4.7.1 Station Servica Water Pump Structure              50 to 104-                      9.4.8.1 I            In  summary,    instrument      sensing    line  freezing  and      fluid precipitation for the System 80+ Standard Design is addressed during plant design, by assuring that-all safety-related systems and components are. enclosed in environmentally controllod buildings wherein the ambient air temperature is maintained above                              .
that which    is  necessary      to    assure  adequate protection.              In addition, the. guidance identified in the applicable SRP Sections, Regulatory ' Guido    1.151,    and the requirements of 10 CFR                50,
          . Appendix A and industry codes and standards including ISA-S67.02 are met as described above. Because the potential for. freezing
.            of or precipitation -in instrument sensing lines is minimir:ed by the design, and the use of insulation and/or heat tracing remains an acceptable alternative, should-it become necessary, this. issue is resolved for the System 80+ Standard Design.
REFERENCES l            1. -  NUREG-0933,  "A Status Report on Unresolved Safety Issues",
L
                  'U.S.' Nuclear Regulatory Commission, April 1989.
: 2. Regulatory Guide 1.151, " Instrument Sensing Lines"              ,  U.S.
j                  Nuclear Regulatory Commission, July 1983.
: 3. NUREG-0800, " Standard Review Plan for the Review of Safety
.                  Analysis Reports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
: 4. 10-CFR 50 Appendix A, " General Design Criteria", office of
;.                .The  Federal    Register,      National  Archives  and      Records Administration.
Amendment F A-22                  December 15, 1989
 
                +'d n
:;p; CESSARitmincanom                                                            j
    .r 1
: 5. ISA-67.02,.' Nuclear-Safety-Related Instrument- Sensing- Line-Piping, and' Tubing' Standards    for. Use- in Nuclear Power                  i
                          = Plants", Instrument Society of America.                                .l i
:i l
o)
J h
                                                                                                    -f .
                                                                                            'i
                                                                                                -- .),
                                                                          .                .~ . .
.-                                                                                            )
t'f 4$ "
;y-f :.
Amendment F A-23                December 15, 1989
 
1 CESSAR !!!ho, 048    LOO FOR CLASS 1E VITAL INSTRUMENT BUSES IN OPERATING REACTORS ISSUE Generic Sefety Issue- (GSI) 048 in NUREG-0933 (Reference 1),
concerns tha' availability of the Class 1E, 120 VAC instrument-bu us a. . 'h- 1 : associated inverters.        The absence of adequate re341reunca      et    some  operating    facilities  jeopardizes- the av.iilability of the instrument, buses and inverters to perform
; ,  their intended safety function, i.e., to provide reliable power L    to safety-related systems ad ' u ponents.
L    ,  review of operating reactre rechnical specifications revealed that some oporating facilit? 'z:hnical specifications did not include limiting conditions ic .: operation (LCO's) for the Class 15, 120 VAC inctrument buses and uss ciated inverters.                      _
    . he Class -1E,    120 VAC instru;ent buses and inverters provide power to safety related equipment in the event of an emergency.
Should these buwes and inverters be out-of-service for an extended period.of time, safety related components or systems may
    .be' unable to perform their intended safety function.                This represents a potential safety concern.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 048 is that the design and operation of the Class 1E vital instrument buses, inverters,      and uninterruptable power supplies (UPS's) shall L    provida-    sufficient      assurance    that    associated    batteries, L    inverters,      battery chargers and regulated transformers are available to support safety systems operation, (i.e., thcre should be a technical specification for the operability of all components which make up the UPS).
Specifically, explicit requirements shall be identified in the form of LCO's for the Class 1E,            120 VAC instrument buses and L    associated inverters.
RESOLUTION Technical Specifications are established and utilized to provide "or the safe operation, inspection, and maintenance of a nuclear facility's systems and components (se's CESSAR-DC Chapter 16).
System      80+    Standard    Design    incorporates    a  variety  of safety-related systems and components which are governed by the Standard Technical Specifications.
Amendment F A-24                  December 15, 1989
 
4 l
C E S S A R n! #,c m .                                                          !
l
                                                                                  .l 1
Among these systems and components is .he onsite Power System                  '
which contains the class 1E, 120 VAC vital instrument buses and inverters (see CESSAR-DC, Chapter 8, Section 8.3).        This system supplies uninterruptable power for safety-related equipment and components as described in CESSAR-DC, Section 8.3.
The LCO of the Onsite Power System, including the Class 1E. Vital bus inverters and will be identified in CESSAR-DC, Chapter 16.                  .
The Technical Specifications will include specific requirements regardin'g plant operational restrictions as they apply to the 120 VAC vital instrument bus and inverters.- operational restrictions are provided in the Technical Specifications to assure the onsite Power System availability and thus an uninterruptable            power source for safety-related systems and components. Incorporated in these restrictions is a periodic evaluation of the Onsite Power System bus condition which considers such availability items as proper breaker and bus alignment, and' adequate bus voltage.
                                                                                ~
In- summary,    the availability of- the .onsite Power. System,                <
including < its    integral  120 VAC vital    bus and associated inverters, is                by the system - design and the Technical Specifications. governed Since the - System is designed to pro, vide an uninterruptable source of power and the Technical Specifications assure. continued availability          during system  operation    as described above, this issue will, therefore, be-closed out for the System 80+ Standard Design upon satisfactory completion of NRC review of CESSAR-DC, Chapter 16.
REFERENCES l  1.      NUREG-0933,  "A Status Report on Unresolved Safety Issues",.
U.S. Nuclear Regulatory Commission, April 1989.
i l
l~
Amendment F A-25                December 15, 1989
 
i 1l CESSAR !!n% m.
049:    INTERLOCKS AND LCOs FOR CLASS 1E TIE BRE7KERS ISSUE Generic Safety Issue (GSI)                    049 in NUREG-0933,                                              (Reference 1),
        . addresser 'the possibility of a compromise to the maintenance of isolation between the 4160 volt Class 1E buses.
This issue is specifically concerned with the use of only one tie breaker to electrically isolate redundant Class ~1E buses. The problem with having one tie breaker separv.ing buses is that isolation between redundant supplies is compromised upon closure due to inadvertent operator action or equipment failure.                                                                                    ,
I~
An additional. issue was raised in NUREG-0933 concerning the use of. interlocks. As described in NUREG-0933, interlocks should be provided to . prevent the unintentional connection of redundant buses =or emergency mources.
ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI 049 are 'that the electrical systems, including the bus tie breakers shall meet the intent of-the guidance identified in Regulatory Guide 1.6 L          (Reference 2) and IEEE Standard 308 in order to neat the requirements of 10 CFR 50 Appendix A (GDC 17), (Reference 3).
In  addition,      these  systems                                  shall                        be  consistent    with  the recommendations identified in NUREG-0933 which include:
: 1. Provide mora than one tie breaker for each cross-connect between recundant Class 1E buses
: 2. If there is only one tie breaker between redundant buses, use a ~ tie breaker only during shutdown (when absolutely necessary)
: 3. Physically ~ disengage            each                                                    tie  breaker  and    " rack-out" L
l~              (withdraw) the breaker fellowing each use
: 4.    " Red Tag" the breaker enclosure to require the breaker to be I_.            kept open L        5. Incorporate Quality Assurance (QA) procedures to reconfirm L
that all breakers are " racked-out" and " Red Tagged" prior to each plant startup L                                                                                                                                                        !
t,                                                                                                                                                      1 l
l-                                                                                                                      Amendment F A-26                                                      December 15, 1989
 
1 i
                                .CESSAR nninc.m.
RESOLUTION                                                                                                            j The-System 80+ Standard Design as described in CESSAR-DC, Chapter 8.0 does not have direct manual or automatic ties between the two Class 1E 4160. VAC power systems. Also, as shown in CESSAR-DC double breakers are provide to maintain independence between the Class-1E and the Permanent Non-Safety 4160 VAC buses.
These breakers          are ~ provided  for abnormal                  scenarMs    such        as Loss-of-offsite Power .and Station Blackout when it is necessary to isolate the Division I & II 4160 VAC buses from the Permanent Non-safety buses. Double breakers (normally open) are also provided for the 4160 VAC standby transformer which feeds each of the Class 1E buses to maintain independence when they are not                                                        i being used during maintenance.                No single failure can prevent operation of. the minimum number of reqttired safety loads. See CESSAR-DC,    Sections - 8. 3.1. 2.1,      8.3.1.2.3              and 8.3.1.2.5 for a discussion of- compliance.            Operating                  and Quality Assurance-                    .". , "
L                                        procedures . governing the engagement / disengagement of the tie l                                        breakers are the responsibility of the Owner-operator.
b,                                      The . electrical systems meet the intent of the*. guidelines identified in Regulatory Guide 1. 6 and IEEE ~ Standard 308. As required by 10 CFR 50, Appendix A (GDC 17), the desi                                        the-power systems provides independence and redundancyensure                          to *gn an    of available source of power to the Engineered Safety                                    Feature systems. Since the guidance and requirements are met, this issue                                              ;
is resolved.for the System 80+ Standard Design.
REFERENCES l
: 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. Regulatory Guide          1.6,    " Independence                Between Redundant l
Standby        (Onsite)    Power      Sources                and    Between        Their l
i-                                              Distribution Systems", U.S. Nuclear Regulatory Commission.
: 3. 10 CFR 50 Appendix A, " General Design Criteria", Offica of The    Federal Register, National Archives and Records Administration.
Amendment F A-27                                December 15, 1989 l
L
_. . _ _ _ _ _ _ _ _ ~ _ _ . . _ _                _ -,                      ..  , - . -.. .-. .              ..  .- -. --
 
l 1
CESSARiniinenio,.                                                                                                                                                                                          J l
RESOLUTION The' System 80+ Standard Design as described in CESSAR-DC, Chapter
                -8.0 does not have direct manual or automatic ties between the two                                                                                                                                              ,
Class 1E 4160 VAC power systems. Also, as shown in CESSAR-DC                                                                                                                                            e double breakers are provide to maintain independance between the Class 1E and the Permanent Non-Safety.4160 VAC buses..
These breakers                                                                        are provided          for abnormal' scenarios such                        as Lo s-Of-Offsite Power and Station Blackout - when it . is necessary to isolate the Division I & II 4160 VAC buses from the Permanent Non-safety                                                    buses.' Double                          breakers- (normally              open)    are      also provided for the 4160 VAC standby transformer which feeds each of the. Class 1E buses to maintain independence when they are not
                ' being used during maintenance.                                                                                    No single failure can prevent                                                              !
operation of the minimum number of. required safety loads. See C1;SSAR-DC, Sections 8.3.1.2.1, 8.3.1.2.3 and 8. 3.1. 2. 5 for a discussion of compliance.                                                                                Operating and Quality Assurance                                                        ~,y.,
procedures governing the engagement / disengagement of the tie l                    breakers are the responsibility of the Owner-operator.                                                                                                                                                    .
l l' ,
The electrical systems meet _ the intent' of the* guidelines                                                                                                                                            ,
identified in Regulatory Guide 1.6 and IEEE Standard 308. As l-                    required by- 10 CFR 50, Appendix A (GDC 17), the desi                                                                                                      the l-                    power systems provides independence and redundancy to ensure                                                                                    ~gn of an-available source                                                                          of power to the Engineered Safety                        Feature                                          .
systems. Since the guidance and. requirements are met, this issue                                                                                                                                    [
is resolved for the System 80+ Standard Design..
REFERENCES 1.-                      NUREG-0933,                                                    "A Status Report on Unresolved Safety Issuev.,                                                                1 U.S. Nuclear Regulatory Commission, April 1989.
: 2.                      Regulatory Guide                                                        1.6,      " Independence            Between Redundant l                                                                                                                                                and            Between        Their Standby                                              (Onsite)            -Power-        Sources Distribution Systems", U.S. Nuclear Regulatory Commission.
: 3.                      10 CFR 50 Appendix A, " General Design Criteria", Office Of The                        Federal                                Register,      National Archives and Records Administration.
Amendment F A-27                            December 15, 1989 4
_ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ . _ _ _ _              _      ~ ___ _ _      , - . -      ... .    ._.      - _ , _    , , _ , . . . . . _ , . _ , -
 
      ,N.
T 1[U
    ~
CESSAR !! nam.                                                                                  -
051:  IMPROVING THE RELIABILITY OF OPEN CYCLE                        ,
SERVICE WATER SYSTEMS ISSUE Generic Safety Issue- (GSI) 051 in NUREG-0933 (Reference 1),
identifies the susceptibility of-the Station Service Water System-
              -(SSWS) to fouling which leads tc plant shutdowns and reduced power operation for repairs.
The SSWS cools the Component Cooling Water System (CCWS) through the component Cooling Water Heat Exchangers and rejects the heat to the ultimate heat sink during normal, transient, and accident conditions. ' The - CCWS in turn provides - cooling water to those safety-related compcnents necassary to achieve a safe reactor shutdown, as well as to various non-safety reactor auxiliary components,                                                                              n ACCEPTANCE CRITERIA The - accepta'nce criterion for the resolution cf GSI                      051 is that the - design of the SSWS shall be cuch that the potential
        .    ' for . fouling of the- piping and heat exchangers be minimized.                              .
This minimize. tion is achievable by: '(1) reducing the number of components which t re directly cooled by the SSWS; (2) employing site-specific          corrosion-resistant        materials and  filtration
              . systems which are acnsistent with the site water chemistry and-                              .s tr.-tment; (3) use af het exchangers with an enhanced thermal ma;; gin.
RESOLUTION The System 80+            Stancard Design SSWS and CCWS are described j              in CESSAR-DC, Sections 9.2.1 and 9.2.2, respectively. The SSWS is l
designed to , serve one Nuclear Steam Supply System (NSSS), and                                  I each NSSS on a multi-unit site will have its own SSWS.
The System 80+ Standard Design features a SSWS which cools only                                  i the CCWS heat exchangers.              Thus the number of components and                        l amount of piping that can become fouled is minimized (Jee                                        I CESSAR-DC,          Section 9.2.1.2).          The CCWS is utilized as an                      1
              ' intermediate system between the SSWS and the safety-related and                                i l              other components being cooled (see Figure 9.2.2-1) .              The CCWS is filled        with  demineralized water and treated with corrosion l              inhibitors. Water quality design parameters applicable to the l'            CCWS are given in Table 9.2.2-1.
l L
l l'
l Amendment F A-28                  December 15, 1989
          --            -. -          _      -.-    _ - . _-    _,      __.    -      __  . . _    _  D
 
CESSAR E%me.
To minimize fouling ' of the CCWS' heat ' exchangers and the SSWS                                  <
piping, prevent. flow blockage and facilitate the maintenance of c lean conditions,      the. following SSWS    design features                    are provided, or required:
              =The SSWS pump structures r.ust be equipped with safety-grade traveling screens vtch a= screen wash system. The screen mesh
              .si e must prevent flow blockage of , the pump inletc, and limit ingestion of biofouling, organics, _ and debris.- (see CESSAR-DC, Section 9.2.1.2.1.4).-                                          ,
      -      Strainers are provided at the SSWS- pump discharges.                        The            .'
strainers are of the automatic backwash type and- are l-            designed to retain particles consistent with the fouling l              design limits of the component cooling water heat exchangers L              (see CESSAR-DC, Section 9.2.1.2.1.5).
      -      . When required by the site-specific. water chemistry - and                              ,
[              environmental regulations, the ultimate heat sink water must l-            be chemically treated to reduce organic and non-organic fouling, corrosion,, scaling, and to keep mud and silt in suspension.  (sas CESSAR-DC Section 9.2.5.2).
L      -      The CCWS heat exchangers are either of the tube and shell or plate and' frame design, dependent upon site selection (see CESSAR-DC Section- 7.2.2.2.1.1). SSWS water flow.is-through                            ;
the tube side of CCWS shell and tube heat sxchangers and at                          '
* l:            a lower pressure than the CCWS to prevent contamination of                            n the. CCWS by in-leakage of SLS water.        In addition, . the                      n nominal flow conditions in CCWS heat exchanger tubes are in accordance with Heat Exchanger Institute standards for power f              plant' heat exchangers.
L
      -      . Adequate tube pull space is provided for periodic tube g              cleaning of the straight tube type CCWS heat exchangers.
The CCWS heat exchengers have a 15 percent thermal performance margin to allow for potential fouling between                                      !
cleaning operations (see CESSAR-DC, Section        9.2.2.2.1.1).
The thermal performance can be verified using temporary l
instrumentation at test connections provided on each heat l
exchanger (see CESSAR-DC, Sections 9.2.1.5 and 9.2.2.5).
      -      Wetted surfaces of the SSWS and CCWS are of materials selected on a site-specific basis to be compatible with the respective cooling water chemistries and water treatments.
The guidelines used for the selection of CCWS heat exchanger
!              tube and tubesheet materials are given in CESSAR-DC Section 9.2.2.2.1.1.
l Amendment F A-29                  December 15, 1989
: 1. . .                                                            -. ,- - . - - - ,.                    -
 
CSSSAR Mne.no.                                    .-
i-                                                                                                                                          +
Sites at-which ice. formation of the ultimate heat sink could E            occur-are to be analyzed to show that the function of the ultimate nnat- sink will not be impaired during winter months. Where required, the, intake structures must be provided "with a means of de-icing,                        such as warm water recirculation, to prevent flow blockage of ' the SMS pump inlets (see CESSAR-DC Section 9.2.5.2).
As" described above, the-System 80+ Standard Decign SSW3 and'CCWS include many design features which minimize the problems that certain plants have experienced with open - cycle service water system fouling or flow blockage due to mud, silt,. ice, corrosion products or aquatic. bivalves. Therefore,                    this issue is resolved for the System 80+ Standard Design.
REFf.RENCES
      '1.. NUREG-0933, "A Status Report on Unresolved Safety Issues",
            =U.S. Nuclear Regulatory Commission, April 1989.                                                                    !          ,
l l
Amendment F A-30                                    December 15, 1989                                      j l
1
__ _.        .. , _ . . . ..      __ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____________i
 
. :CESSAR W Lu..
0571 EFFECTS OF FIRE PltOTECTION SYSTEMS ACTUATION ON SAFETY-RELATED EOUIPMENT ISSUE Generic. Safety Issue (GSI), 057 in NUREG-0933 (Reference 1) , .
addresses the potential for safety-related equipment to become                                                          4 inoperabla' because of ' water spray from the- fire protection                                                          i system..      IE information Notice 83-41 (Reference 2) identified I
expsriences in - which actuation            of fire suppression systems                                  -
caused damage to-safety-related equipment.
ACCEPTANCE CRITERIA The acceptanca criteria for the resolution of USI 057, is thet the fire protection system - be designed to preclude damagiries safety-related equipment and rendering the equipmenc' inoperable.                                                  u In addition, the fire protection system shall be designed to meet                                                  i 10 CFR 50 Appendix A (GDC 3) (Reference 3); which states in part:
    " Fire detection and fighting systems of appropriate capacity and                                                      ;
capability shall be provided and designed to minimize the adverse effects of ? ires on structures, systems, and components important to safety.        Fire fighting systems shall be designed to assure that      their    rupture  or    inadvertent                    operation does not significantly- impair the safety capability - of those structures, systems, and components."
  ~ RESOLUTION The ~ System 80+ Standard Design is designed to preclude water npray from the fire' protection system onto safety-related equipment. T1.e sprinkler systems proteccing the safety-related equipment is of the automatic sprinkler ty s.                        Actua*. ton of these sprinkler systems requires the            opening cf the fusible                            link sprinkler heads and detection by combustible-products and/or heat detectors.      In addition,    the operator has the capability of isolating flow from the control room by isolating the Sub-sphere Building headers or, locally by manual isolation valves.
In order to prevent damage due to flooding, upon actuation of sprinkler systems, floor drains are erovided and equipment                                    is located to preclude the flooding of the equipment. In addition,
  .in order to further reduce potential damage to safety-related equipment upon actuation of sprinkler systems, ' equipment is shielded and conduit ends are sealed where required based on interaction      reviews  during    detailed              design        and        as built walk-downs.
Amendment F A-31                                  December 15, 1989
 
CESSAR !!airicam.
i systems are designed to preclude      l      d
      .Since. the Fire. Protection inadvertant'actaution        and thus minimize      damage designed  in accordanceto safety-re ate            l equipment and because -these systems arethis issue is resolved for the                              ,
with 10 CFR 50 Appendix A (GDC.3)-.
System 80+ Standard Design.
pzrmazucas "A Status Report on Unresolved Safety. Issues",
NUREG-0933,
: 1. U.S. Nuclear Regulatory Commission, April.1989.                          i 2,
I E Information Notice 83-41; " Actuation of Fire Suppress      i        t";
on          y System Causing Inoperability cf Safety-related Equ pmen SSINS No. 6835.
p                                        " General Design Criteria", Office of    Records l
10-CFR 50 Appendix  A,                Archives      and                                  !
: 3.                    Register,    National the  Federal                                                                    .
                -Administration.                                                                  J l-.
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Arendment F December 15, 1989 A-32                                                        .
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g CESSAR M* nema 9,$1L IDENTIFICATION OF PROTECTION SYSTEM INSTRUMENT BEMSING LINES t
ISSUE Generic Safety - Issue (GSI) 064 in NUREG-0933 (Reference 1) ,
addresses the establishment of guidance for the identification of the mechanical sensing lines which interface with safety-related instrumentation and controls systems.
Sensing    lhes are an integral part of the safety-related (protectier.)  systems,  and are essential to their reliable operation. Therefore, identification of these lines facilitates verification that - these lines are appropriately separated and protected.
Industry    has also developed a standard for safety-related instrument sensing lines, ISA-S67.02-1980 (Reference 2), which includes identification criteria. As part of establishing- its guidance for safety-related instrument sensing lines, the NRC endorsed ISA-S57.02-1980 in Regulatory Guide 1.151 (Reference 3).
ACCEPT &MCE CRITERIA                    ,
The acceptance criteria for the resolution of GSI 064, are that sensing lines which interface with safety-related instrumentation and controls shall be identified in accordance with ISA-S67-02
        -1980 and meet;the intent of the guidance provided in Regulatory Guide 1.151.
Specifically,    the  instrumentation    sensing lines shall meet Section 5.3 of the ISA Standard.        Section 5.3, in part, states that the instrument sensing lines related to safety-related instrumentation will be identified and color coded.
RESOLUTION The    System    80+    Standard  Design    includes  safety-related instrumentation and controls which use mechanical sensing lines.
Thase sensing lines are identified and color coded in accordance with Regulatory Guido 1.151 (see CESSAR-DC Section 7.1.2.31).
In addition, the guidance identified in Regulatory GuLles 1.151 and 1 ~. 7 5 is imposed as design criteria for the routing of 1E (safety-related) and associated cabling and sensing lines from sensors.
Since the System 80+ safety-related inscrumentation and controls (including the sensing lines) meet the critaria given in ISA-S67.
d;    02-1980 as invoked by the guidance given in Regulatory Guide 1.151, this inreue is resolved for the System 80+ Standard Design.
Amendment F A-33                Decerber 15, 1389
 
CESSARiniincuior                                                              i REFERENCES
: 1.      NUREG-0933,  "A Status Rept,rt on Unresolved Safety Issues",        '
U.S. Nuclear Regulatory Commission, April-1989.
            '2.      . Standard ISA-S67-02-1980, " Nuclear. Safety-Related Instrument Sensing Line Piping And Tubing Standards For Use In Nuclear Power Plants",. Instrument Society of - America, February 17 , .
1943.-
: 3.      Regulatory Guide -'1.151, " Instrument Sensing      Lines",  U.S.
Nuclear Regulatory Commission, July 1983.
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Amendt:ent F A-34                  December 15, 1989 l
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    .CESSAR !!ninemon i
066i    STEAM GENERATOR REOUIREMENTS ISSUE Generic Safety Issue        (GSI)  066 in NUREG-0933          (Reference      1),
addresses the potential for and the safety implications of steam generater tube ruptures (SGTR) in PWR's. Unplanned radioactive affluent releases to the environment and loss of primary coolant
      . inventory as a result of a SGTR are also addressed.
      -Plant operating ' experience has demonstrated that a number of problems which have arisen with PWR steam generator tubes have resulted in steam . generator tube degradation leading to leaks I      and/or ruptures. To date, different forms of steam generator tube degradation have been identified,          including: stress corrosion L      cracking,    wastage,    intergranular    attack,    denting,      erosion-l      corrosion, corrosion cracking, pitting, fretting, support plate degradation, and mechanical wear (e.g., vibration fretting) as described in'NUREG-0844 (Reference 2).
[- ,
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 066 is that the secondary system including the steam generators and condenser shall be designed, manufactured, tested, inspected, and operated in accordance with accepted industry codes and standards.
The steam generators shall meet the requirements of Sections III-                    9 and M of the ASME B&PV Code for design, manufacture, test, and inspection. Also, steam generator design shall meet the intent of the guidance-given in SRP Sections 5.4.2.1, Rev. 2 and 5. 4. 2. 2, Rev. 1 (Reference 3) for steam generator materials, quality assurance, inservice tube inspection, and secondary side water chemistry.
RESOLUTION The System 80+ Standard. Design specifies a variety of accepted industry codes and standards to assure the integrity of both the steam-generators and the main condenser.
In  addition,  stringent    secondary    condensate        and  feedwater che :istry requirements are employed to maintain steam generator and condenser integrity during operation.
The steam generators and condensers are designed to meet accepted industry codes and standards including the specific requirements cf the ASME B&PV Code Sections III Lnd XI and the Heat Exchanger Institute Standards as identified in CESSAR-DC Sections 5.4.2 and 10.4.1.2.
Amendment F A-35                      December 15, 1989
 
i CESSAR inancuion                                                            I l
CESSAR-DC, Section 5.4.2.4.1 describes the materials used for the steam generator' tubes. CESSAR-DC, Section 5.4.2.4 describes the materials for the remainder of the generator components. Steam generator- test and inspection criteria            are discussed in          ,
CESSAR-DC, Section 5.4.2.5. Also, steam generator secondary side              '
water chemistry is discussed in CESSAR-DC, Section 10.3.5.
In summary, the secondary system, including the steam generators and condenser, meets the requirements specified in accepted industry codes, and standards. In addition, steam generator materials and secondary side water chemistry meet the intent of the guidance given in SRP Section 5.4.2.        Therefore, this issue is resolved for the System 80+ Standard Design.
REFERENCES J
I-  1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
l U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0844, "NRC . Integrated Program for the Resolution of Unresolved Safety Issues A-3, ,A-4, and A-5 Regarding Steam Generator    Tube Integrity",      U. S. Nuclear    Regulatory Commission, September 1988.
: 3. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports. for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
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Amendment F A-36                    December 15, 1989
 
          ,2-li; CESSAR1!Nficario 079:  UN1MiLYEED REACTOR VESSEL THERMAL STRESS                                            l j
DURING MATUREL CONVECTION COOLDOWN                                                  !
I l-
!            ISSUE Generic , Safety Issue    (GSI) 079 in NUREG-0933 (Reference 1),
jj            identifies the potential for the stresses in the reactor vessel flange area or studs to exceed the allowable during its design.
L            lifetime because    of- a  previously_ unanalyzed thermal                stress                  1 introduced by the natural convection cooldown event.                                                :
A natural convection cooldown event occurred _at the St. Lucie 1 nuclear power generating _ station. During the course of this L
l' event, steam voiding occurred in the reactor vessol head area.
j            Upon analysis,    concern .was raised over previously unanalyzed'
!              reactor vessel  thermal' stresses.      The concern focused on the
                                                                                                          +
L possible existence-of an axial temperature gradient of 150 to 200 degrees F in the vessel flange and studs.                                                      m L
l            -The safety concern arises because this event could produce
"              thermal stresses in the flange , area or in the studs that may exceed'the ASME B&PV, Section III Code (Re*ference 2) allowables when added to the stresses already . considered.                  Moreover,.the l              cafcling of these temperature gradients over the life of the plcnt l              has the potential to-cause a reduction in the fatigue margin of                                ',
the vessel.
ACCEPTANCE CRITERIA                                                                              ,
The acceptance criterion for the resolution of GSI 079 is that the design of the reactor pressure vessel (including the head and-studs) shall accommodate the thermal stresses caused by a natural convection cooldown event. These thermal stresses, when added to stresses from events that are presently analyzed, shall not exceed the stress limits specified in the ASME B&PV Code, Section                                ,
III.
RESQLUTION Stress analyses were performed to determine the effects of a natural circulation cooldown event (similar to that of the St.
Lucie occurrence) on both the St Lucie " class"              reactor vessel and The analyses concluded the System 80 " class" reactor vessel.
that should natural circulation cooldown of the reactor coolant system be required and should vessel head voiding subsequently occur,    the resulting thermal stresses would not. cause any thermal, hydraulic, or fatigue damage to the reactor vessel and its integral components over their design lifetime.
Amendment F A-37                        December 15, 1989
 
CESSAR Mincan.
Furthermore, the System 80+ reactor vessel, which is designed to the ASME B&PV Code, Section III (see CESSAR-DC, Section 5.3), is essentially  identical. to      the    System 80  reactor  vessel.
Specifically, the vessels have the same material composition and overall- dimensions -and are of similar geometry (with the-exception of the direct vessel injection nozzles) as described in            1 CESSAR-DC, Table 1.3-1, and Figure 3.9-9.        Because the reactor vessels for both " classes" of plants are virtually the same and since the stress analyses consider the materials,' dimensions and geometry of the vessel, the analyses performed subsequent to the            i St. Lucie 1 avant apply to the System 80+ reactor vessel.-                  -
In summary, the addition of the dynamic, thermal and fatigue effects of a natural convection cooldown on the System ~ 8 0+
reactor vessel does not result- in the vessel stresses or fatigue usage factor exceeding the allowable limits specified in the-ASME              !
B&PV Code, Section III. Therefore, this issue is resolved for-the System 80+ Standard Design.
REFERENCES                                                                    ;
: 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",            _!
U.S. Nuclear Regulatory Commission, April.1989.                          !
: 2. American Society of Mechanical Engineers, Boiler & Pressure
      = Vessel _ Code, Section III (Nuclear).
Amendment F A-38                December 15, 1989
 
I
  ,                  - CESSAR !nW,emo,,
:                                                                                                                                  l 0821      BEYOND DESIGN BASIS ACCIDENTS IN SPENT FUEL POOLS.
ISSUE Generic Safety Issue (GSI) 82 in NUREG-0933 (Reference 1),
addresses the potential for a beyond-design-basis accident in which the water is drained out of the spent fuel pool, allowing                                          ;
the Zirealoy fuel cladding to ignite and thus- release fission
                      . products from the spent fuel to the atmosphere. (The spent fuel                                          i pool is.usually located outside the primary containment.)-
The risk of beyond-design-basis accidents                        in              spent  fuel
                      - pools was examined in WASH-1400 (Reference                  2),                  when it was concluded that the risk was orders of magnitude below those involving the reactor core.            The. issue has been re-examined
    .                    by the NRC because of subsequent developments: the' storage of-spent fuel at reactor sites in high . density racks (instead of reprocessing); and laboratory studies indicating the possibility of    fire propagation        between assemblies in an air cooled environment.        The two developments together provided the basis for a hypothesized accident scenario not previously considered.                                  .
After fur.ther NRC evaluation          it was concluded that further reduction in the already very low risk from the spent fuel pool I
accident would still leave a comparable risk due to core damage
!                        accidents,'and therefore, no additional requirements for the safe storage of spent fuel in the primary spent fuel storage pool are warranted.        This resolution by the NRC assumes that all current applicable        requirements and guidance have been met.                                The
                        ' Regulatory Analysis for the resolution of GSI 082 is documented in NUREG-1353 (Reference 3).
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI-82 is that the design of the spent fuel pool, storage racks, water cleanup and cooling system, and load handling equipment in the spent fuel pool area.shall meet applicable current requirements, consistent with the guidance of the Standard Review plan (SRP) Sections 9.1.2 through 9.1.5 and' Regulatory Guide 1.13 (References 4 and 5 respectively).        (Acceptance criteria for that portion of the fuel handling system that handles heavy loads (see SRP 9.1. 5) , i.e.,
loads which exceed the combined weight of a fuel assembly and its handling device, are also provided in conjunction with Unresolved Safety Issue        (USI)  A-36,  " Control of Heavy Loads Near Spent Fuel".)
Amendment F A-39    -                          December 15, 1989
_ _ u _ ____ _ _ _ ____. _ _ _ _ _ . .
 
i CESSAR nn*nemo,.
e R189LUTION The System 80+ Standard Design includes a spent fuel wet storage facility, together with its associated handling systems, that meets the intent of Regulatory Guide 1.13 (Reference 5) and conforms to the relevant requirements of GDC 2,              4,      5, 44, 45. 46, 61, 62 and 63 (Reference 6).
The spent - fuel pool and the storage racks are described in CESSAR-DC Section 9.1.2, _.the spent fuel pool cooling and cleanup system is described in Section 9.1.3,                    and the fuel handling system (which includes the equipment for handling heavy loads) is described in Section 9.1.4.          (A more complete description of how the fuel handling system conforms to acceptance criteria for the handling.of heavy loads is provided in the response to USI A-36.)
Since the acceptance criteria are met for the spent fuel storage facility, this issue is resolved for the System 80+ Standard                            ,
Design.                                                                                .
REFERENCES                    ,
: 1. NUREG-0933, "A- Status Report on Unresolved Safety                      Issues",'
L        U.S., Nuclear Regulatory Commission, April 1989.
!  2. WASH-1400, " Reactor Safety Study - An Assessment of Accident l        Risk in U.S. Commercial Nuclear Power Plants", U.S. Nuclear Regulatory Commission, October 1975.
: 3. NUREG-1353,    " Regulatory Analysis- for the Resolution of Generic Issue 82, 'Beyond Design Basis Accidents in Spent Fuel Pools' " ,- U.S. Nuclear Regulatory Commission, April 1989.
: 4. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition",
U.S. Nuclear Regulatory Commission.
: 5. Regulatory Guide 1.13,        " Design Objectives for Light-Water Reactor Spent Fuel Storage                  Facilities at Nuclear Power 3
Stations",  U.S. Nuclear Regulatory Commission, Revision                2, December 1981.
: 6. 10 CFR Part 50 Appendix A, " General Design Criteria for Nuclear Power Plants",              Office of the Federal Register, National Archives and Records Administration".
Amendment F A-40                        December 15, 1989
 
                  ..                                          -.              .        =
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:CESSARun%m.
083:  CONTROL ROOM EABITABILITY                                                                  1 l
l ISSUF Generic Safety Issue (GSI) 083 in NUREG-0933 (Reference 1), deals with ensuring that the control                        room design is adequate                              to preclude the loss of control room habitability following an accidental release' of external airborne toxic or radioactive material or smoke which could impair the control room operators'                                                                l ability to safely control the reactor.                                                                                          !
1    ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 83 is to                                                                      ;
verify that the control room design is adequate to prevent the                                                                  I loss of habitability of the control room during an accident.                                              The                  .
design must meet the intent of the guidance given in Standard                                                              v    !
Review Plan (SRP) Sections 6.4 Rev. 2 and 9.4.1 Rev. 2 (Reference                                                              -l 2). The design must be in accordance with 10 CFR 50, Appendix A, General Design Criteria (GDC) 2, 4, 5, 19, and 60 (Reference 3).                                                                !
1 RESOLUTION The System 80+ Standard Design main control room Habitability System is- described in CESSAR-DC Sections 6.4 and 9.4.1 and the                                                            .
design bases are given. in Section 6.4.1. The control room is a structure which is important to safety and as such is designed to                                                          4-withstand- the              effects    of    natural          phenomena                (earthquakes,.                        l hurricanes, etc.) and postulated accidents . and missiles. The                                                                  l design is, therefore, specifically in accordance with GDC 2 and 4-                                                              .
(see CESSAR-DC,              Secticns  3.1.2      and  3.1.4).                The System              80+-                  l Standard Design is based ^ on non-shared systems (see CESSAR-DC, Section 1.2.1.3) and therefore GDC 5 is met (see CESSAR-DC, Section 3.1.5).              The design of the control room permits safe occupancy during abnormal conditions and meets the requirements of GDC 19 (see CESSAR-DC, Section 3.1.15) .
The Control Room Ventilatic,n and Air Conditioning Systems are l'
designed for uninterrupted safe occupancy of the control room during post accident shutdown in accordance with-GDC 2, 4, 5, 19 and 60 (see CESSAR-DC, Section 9.4.1.1). Fire protection for the                                                                i L
control room is provided by alarn systems and portable fire extinguishers            (see CESSAR-DC, Section 6.4.1).                                The        testing l:
i      requirements for the Habitability System are                                          identified in CESSAR-DC Sections 6.4.4 and 9.4.1 4.
l-
: l.                                                                                                                                    l L                                                                                                                                      ;
Amendment F A-41                                  December 15, 1989 1
a    - - . ,    . - .- - .            ~    . __.      . _ _ _ _ _ _ _  _ , _      _    _ _ , _ _        _ _ _ _ . ,
 
CESSAR !!nir .rio.
t Since the. control room design prevents the loss of control room habitability during accident conditions, and since all of the NRC requirements and guidance are met, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. . NUREG-0933,  "A - Status - Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission,; April 1989.
2.-  NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
3.. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear' Power Plants",  Office of The Federal Register, National Archives and Records Administration.
T L
Amendment F A-42                December 15, 1989
 
1 o  CESSAR !!n%m.
l 0932  STEAM BINDING OF AUZILIARY FEEDWATER PUMES, ISSUE                                                                                                  I Generic Safety Issue (GSI) 093 in NUREG-0933 (Reference 1),                                            '
addresses the potential for a common mode failure of the Auxiliary or Emergency Feedwater (ETW) System resulting from steam binding of the EFW pumps caused by back leakage of main feedwater through check valves.                              The EFW system is used to provide water to the steam generators in the event of a loss of the Main Feedwater (MFW) System.
The EFW system may be isolated from the MFW system by a check valve or one      or more                    isolation valves      (depending upon the specific design) to preclude hot main feedwater from entering the EFW system. However, operating experience has shown that check valves tend to leak, thus permitting het main feedwater to enter the EFW . system. This hot feedwater can subsequently flash to steam in the EFW pumps and discharge lines resulting in steam binding of the pumps.
In addition, the EFW piping is sometimes arranged such that each EFW pump is connected through a single check valva (which is used
* to prevent back leakage) to piping which ia common to two or three pumps. This arrangement creates the potential for common mode failures as the hot .feedwater leaks back through the check valves into other EFW pump (s).
Because of the NRC's concerns regarding steam binding a Generic Letter (GL) 88-03 (Reference 2) was issued to the industry and is the final resolution to this issue. The purpose of the letter is to implement monitoring and corrective precedures to minimize the likelihood of steam binding of the EFW system pumps.                            One of the corrective actions to be taken is the monitoring of EFW pump discharge'- piping      temperatures                        to  ensure  that    the      fluid temperatures remain at or near ambient.
I    ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI 093 are that the design of the EFW system shall be such that the potential for steam binding of the EFW pumps is minimized and that the EFW
:. system shall meet the intent of GL 88-03.
1
)    RESOLUTION The EFW system in the System 80+ Standard Design includes two major independent " trains" with each train aligned to supply its                                      ;
!    respective steam generator.
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Amendment F A-43                December 15, 1989 l
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CESSAR M h ..
L                                                                                                                                  j i                                                                                                                                    l q.
Each major train, which consists of two subtrains (see CESSAR-DC, Section 10.4.9), contains:                                                                                                      j
: l.      one emergency feedwater storage tank (EFWST),                                                                          l
: 2.      a motor-driven and a steam-driven pump (each with a capacity                                                          l I
of 500 GPM),-
one flow control valve per subtrain,                                                                                  I 3.
: 4.      one isolation valve per subtrain,
: 5.      one check valve per subtrain,
!    6.      a cavitating venturi, and l  7.      specified instrumentation.
l The main defense against steam binding of the EFW Pup 5 results                                                                l L
from the system design for normal plant operation.                                                                            ,
l Although some plant systems operate with the flow control and the                                                              I
!    isolation valves open during normal - plant operation, the system                                                            j
!    80+ Standard Design EFW system is designed to operate with the                                                          m isolation valves closed.                  The closed isolation valves act as a backup to the EFW line check valves, thus providing redundant isolation of the EFW System ,from the MFW System.                                                CESSAR-DC Section 10.4.9 states that the isolation valve will be closed                                                                  i during normal plant operation.                            When a steam generator low level                                      l L    occurs, the Emergency Feedwater Actuation Signal (EFAS) starts                                                                l l-  the EFW pumps (the motor and steam driven) , opens the isolation l'  valves, and assures that the feedwater flow control valves are open, allowing EFW flow to each steam generator..
i    Each EFW subtrain is separated from the other.                                  Each subtrain has its own suction line from the EFWST, its own discharge line through the steam generator isolation valve and check valve, and the pump crossover lines contain                                redundant,        locked closed isolation valves.        Thus,    the            potential  for common        mode        failure of steam binding of all EFW pumps does not exist, should one set of steam generator isolation and check valves leak.                                    The EFW pump                            i suction and recirculation lines are normally open so that, should                                                            I leakage'of a steam generator isolation and check valve occur, any resulting steam can be vented through the EFWST vent.
;    Associated instrumentation is provided for each train to assure l    adequate control and monitoring of the EFW system.                                    Temperature indicators (TI's)        are located between the flow control and l    motor-operated isolation valves (MOV's).                            These TI's provide a
;    direct indication of the fluid temperature and alarm on high
[    fluid temperature in the EFW system downstream of the EFW pumps.
!'    This alarm warns the operator that leakage through the steam
  . generator isolation valve                              and check valve            is        occurring.                        .
Therefore, these sensors provide indication to the operator of                                                                l the potential of steam binding of the EFW pumps.                                                                              l l
l Amendment F                                  ,
A-44                      December 15, 1989                            l
 
b CESSAR !!!hio                                                                                                  i l        In summary, the System 80+ Standard Design addresses steam binding of the EFW pumps in four ways.                      First, each train is
: l.      equipped with two normally-clor,ed isolation valves namely a Mov and a check valve.            Thus redundant isolation of the EFW system from the main feedwater system and associated steam generator la achieved.      Second, each subtrain is separated from the other such                                          ;
that a leak of a single set of valves does not affect all of the pumps. Third, TI's in each EFW pump discharge line alert the
,        plant operator should valve leakage be present.                      Finally, open                            ,
L      : lines permit valve leakage to be vented through the EFWST vents.
Since the EFW Fystem in the System 80+ Standard Design meets and
      ~ exceeds the intent of GL 88-03, this issue is resolved for the System 80+ Standard Design.
BIZEBEMCXE
: 1. NUREG-0933,        "A Status Report on Unresolved Safety Issues",                                  a U.S. Nuclear Regulatory Commission, April 1989.                                                    ,,-  ,
1
      . 2. Generic Letter 88-03, " Steam Binding of Auxiliary Feedwater                                          I Pumps", U.S. Nuclear Regulatory Commission.
5                                        .
l l
l.
l l
l l
i-l l                                                                          Amendment F A-45                        December 15, 1989
 
CESSARUn bion i
i l
103t. DESIGN FOR PROBABLE MAXIMUM PEEQIPITATION                                    l j
ISSUE 1
Generic Safety Issue (GJI) 103 in NUREG-0933 (Reference 1),                                  )
addresseu the accepted methodology used for determining the design flood level for a particular reactor plant site. Accurate I
determination of the design flood level for a specific reactor
;        alte is necessary in crder to ensure adequate protection of                                  )
safety-related equipment against possit - site flooding.                                    j Reactor plant sites .are designed to s ,commodate maximum flood                              '
level because flooding could disable safety-related equipment.
Historically, estimating design flood levels for specific reactor plant sites has been based upon input data for probable maximum                              .
flood (PMF) provided by the U.S. Army Corp. of Engineers for.the specific site. The guidance identified in the standard Review Plan (SRP) Sections 2.4.2, Rev. 3, and 2.4.3, Rev. 3 (Reference
: 2) is used in predicting design flood levels. Furthermore, general requirements are defined in General Design Criteria (GDC) 2 (Reference 3) .. The SRP's state that " design basis flood levels" incorporate the most seve're historical environmental data with
        " sufficient margin". What is considered to be                          "su f fif.,1 ent margin" and procedures for estimatina PMF's are identified in Regulatory Guides 1.59 and 1.102, and ANSI /ANS 2.8 (Refetences 4, 5, and 6).
ACCEPTANCE CRITERIA The- acceptance criterion for the resolution of GSI 103 is that the site chosen for a commercial nuclear generating facility shall be designed to accommodate a maximum expected flood from precipitation without jeopardizing the safe operation of the facility, in accordance with the guidance given in SRP 2.4.2, Rev. 3 and SRP 2.4.3, Rev. 3. Also, the facility design, including        structures,        systems,    and  components    important      to safety, shall meet the criteria specified in 10 CFR 50 Appendix A 1,
l-(GDC 2).
RESOLUU.Q.H The  System          80+    Standard  Design    is  designed    to    meet    the requiremtnts of GDC 2 as described in CESSAR-DC, Section                        3.1.2.
The System 60+ Standard Design is based upon a set of assumed site-related            parameters. These    parameters  were    selected      to envelope most potential nuclear power plant sites in the United States. A summary of the assumed site design parameters, i
iJ      including maximum flood level, is given in CESSAR-DC, Section
: 2. 0, Table 2. 0-1.
Amendment F l                                                  A-46                      December 15, 1989 l
 
E                                                                                                                l 1
i CESSARnuh -                                                                                                    l 1
I I
i l
i Detailed site characteristics based upon historical site specific                                                i environmental data will be provided by the site owner-operator                                                  l
  -for any specific application. The site owner-operator will review                                              I and evaluate these characteristics to ensure compliance with the enveloping assumptions of Table 2.0-1.
1 Since the System 80+ Standard Design is designed in accordance with GDC 2 for the most severe expected environmental conditions, including flooding, tornado, hurricane etc., and meets the intent of SRP Section 2.4.2, Rev.              3, and SRP Section 2.4.3, Rev. 3 with respect to plant design, therefore this issue is resolved for the System 80+ Standard Design.
REFERENCE 8 1<    NUREG-0933,        "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0800, " Standard Review Plan for the Review of Safety                                          5 Analysis Reports for Nuclear Power Plants-- LWR Edition",
U.S. Nuclear Regulatory Commission.                                                                  "
: 3. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear Power Plants", Office of The Federal Register, National Archives and Records Administration.                                                                  .,
: 4. Regulatory Guide            1.59, " Design Basis Floods for Nuclear Power Plants",          U.S. Nuclear Regulatory Commission, August 1977.
: 5. Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants", U.S. Nuclear Rogulatory Commission, September 1976.
: 6. ANSI /ANS        2.8,    " Standard      for          Determining      Design    Basis Flooding at Power Reactor Sites", American Nuclear Society.
Amendment F A-47                              December 15, 1989                1
 
CESSAR E!Enem.
q l
106  PIPING AND THE USE OF COMBUSTIBLE GASES IN VITAL AREAS                                                                                  1 4
Generic Safety Issue                        (GSI)    106 in NUREG-0933                          (Reference                        1),
addresses            the - issue            of combustible                gases          accumulating                              in buildings containing                        safety-related                equipment.                Except                          for hydrogen, most combustible gases are used in limited quantities and for relatively short periods of time. Hydrogen is stored- in high pressure storage vessels and supplied to various systems in the Auxiliary Systemu Building through small diameter - pipe.                                                                          A l    leak or break in this pipe ' could result in a combustible or l
explosive mixture of air and hydrogen posing a potential-loss of safety-related equipment.
SRP Section              9.5.1,    (Reference 2)                      addresses' this concern for
                                                                                                                                              ~'
plants under construction and new plant designs.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of USI l'06, is that the hydrogen and other combustible piping be designed to preclude large releases and accumulation of combustible or explosive gases in buildings which anclose safety-related equipment.                                                  This can be accomplished either by designing piping to preclude failure or providing means to limit the amount of hydrogen leakage in the event of a pipe rupture. Furthermore, in consideration of the above, the designer shall consider the guidance described in-SRP Section 9.5.1.
RESOLUTION The System 80+ Standard Design incorporates various Compressed Gas Systems as described in CESSAR-DC, Section 9.5.10.                                                                              The compressed gas systems provide a variety of gases (e.g., hydrogen and nitrogen) under pressure, "or numerous                                                  plant operating applications              including            welding,                equipment,            instrumentation, system purging, inerting and diluting.
The systems typically consist of high pressure gas cylinders, pressure regulators and pioing to distribute the gases throughout the plant.              These non-safety-related compressed gas systems are
    - designed. to assure that their failure does not jeopardize the operation of any safety-related system and/or component (see CESSAR-DC, Section 9.5.10.1) .                          Furthermore, with respect to the hydrogen compressed gas system, the system is designed to be isolable and a leak detection system is included.                                                                          Also, Amendment F A-48                                        December 15, 1989
 
CESSAR12%,m,.
in accordance with the guidance given in SRP Section 9.5.1, the piping which transports the- hydrogen within the                                                        containment sub-sphere is either sleeved, includes excess flow shutoff valves or is designed to Seismic Category I in accordance with the guidance          given in SRP Section 9.5.1 for limiting hydrogen accumulation (see CESSAR-DC, Section 9.5.10.2).
Since the non-safety-related compressed gases systems are                                                                              '
designed so that their failure will not jeopardize safety-related equipment, and the hydrogen and other combustible gas piping is designed to preclude large releases and accumulation of combustible or explosive gases in buildings which enclose
    - safety-related - equipment, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933, "A Status Report on Unresolved Safety. Issues",                                                                ,.
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0800, " Standard Review
* Plan- for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
U. S. Nuclear Regulatory Commission.
s a
l l
l 1
Amendment F A-49                              December 15, 1989
 
                                                                                                                                                                                                            ?
CESSAR1!!L m.
l 119.1t            PIPING RUPTURE REOUIREMENTS AND DECOUPLING OF SEISMIC AND LOCA LOADS                                                                                                                                          l 1
ISSUE Generic Safety Issue (GSI) 119.1 in NUREG-0933                                                                                                                (Reference 1),
addresses the recommendations of the NRC Piping Review Committee (PRC)    on how . the NRC should modify their requirements with                                                                                                                                  ,
respect to design loads. on pipes in safety-related systems and                                                                                                                                    I high-energy lines important to safety in. new nuclear power
          -plants. With respect to postulated pipe breaks, the scope covers all high-energy lines.
For    GSI-  119.1,            the          PRC    recommended                                                  the                                utilization                  of            l technology in determining the need to                                                                                                                ;
Leak-Before-Break (LBB) consider the dynamic effects of pipe breaks. The "l:mited scope" rule of General Design Criterion (GDC) 4, which existed at the time of the recommendation, allowed the application of LBB on main coolant loop piping in pressurized water reactors.
Successful application of LBB to a piping system eliminated the need to consider the dynamic effects of breaks in that pipe. 'The PRC . recommended, and the NRC revised GDC 4 to define, a " broad scope" approach allowing the application of LBB to all high energy lines in nuclear power plants.                                                                    Revisions to Standard                                                                    l Review Plan (SRP)                  Sections      3.6.1                      and              3.6.2                                      to implement the                                        l
          " broad scope" rule and to eliminate postulation of arbitrary intermediate pipe breaks were promulgated.
An additional PRC recommendation to decouple SSE and pipe break loads in the mechanical design of components and their supports, which would require a revision to SRP Section 3.9.3, has not yet been accepted or implemented by the NRC.
ACCEPTANCE CSITERIA Ths acceptance criterion for the resolution of GSI 219.1 is that l          the    piping        design              shall      be                  in                accordance                                                            with        those recommendations of the PRC which have been implemented by the NRC l
(GDC 4 and the Standard Review Plan revisions). SRP Section 3.6.2 I          eliminates the requirement to postulate arbitrary intermediate breaks. In addition, SRP Section 3.03, which is currently in draft form, implements the " broad scope"' rule of GDC 4 and endorses the LBB methodology contained in NUREG-1061 Volume 3 (Reference 2).
1 Amendment F A-50                                                                                                    December 15, 1989 l
l
 
m                                                                                    ,
                                                                                      )
CESSAR1!nknor l
l l
RESOLUTION                                                                        !
The design of the piping for the System 80+ Standard Design meets E-  GDC 4 and the guidance of the Standard Review Plan as follows:
: 1. As stated in CESSAR DC, . Section 3.6.2.1.3, - LBB methodology
-        is used to eliminate the postulation of - breaks in the following System 80+ piping systems:
Main Coolant Loop, Surge-Line, Main Steam Line, Safety Injection Line, and shutdown Cooling Line.
l  2. Postulation of arbitrary intermediate pipe breaks in all i        piping systems is eliminated. Postulated break locations are described in CESSAR DC, Section 3.6.2.1.4.                              y
: 3. Piping, component,. Land component support loads from SSE and the pipe breaks remaining after application of LBB continue to be combined in accordance with SRP Section 3.9. 3 s The            '
loads are . combined . on a square root of the sum of the squares basis (CESSAR DC, Section 3.9.3.1 and Tables 3.9-10 through  14). The exception to .this approach is that                7 asymmetric blowdown loads in the reactor vessel, which are
* associated with small break LOCAs that remain, are accounted for by increasing the SSE loads by a small, conservative                  '
factor (CESSAR DC, Section-3.9.2.5).
Loads en building components from SSE and remaining pipe breaks continue to be combined on an absolute sum basis.
Because the design methodology is in accordance with current regulations and NRC guidance, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: i. 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues,"
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-1061, Volume    3,  " Evaluation of Potential for Pipe i        Breaks," U.S. Nuclear Regulatory Commission, November 1984.
i Amendment F A-51              December 15, 1989
_  _    .- -  ~ _ _
 
1 1:
    - CESSAR !!nLm.
119.2t  PIPING DAMPING VALUES l
Issus                                                                                    l l
l    Generic Safety ' Issue (GSI) 119.2 in NUREG-0933              (Reference 1),            l l    addresses the recommendations of the NRC Piping Review Committee                        !
1 (PRC) on how the NRC should modify their requirements for the damping values to be used in the dynamic analysis of nuclear power plant piping systems.
Piping dynamic response would, in general, be more accurately                            l predicted if higher piping. damping values were used than those identified in the - current regulatory guide.      The use of higher damping values results in nuclear plant piping systems having                            .
significantly fewer snubbers - and supports and an overall better                      )
balance of design considering all piping loads. A significant                            I decrease in the number of snubbers and supports allows for better inspection of equipment and components at significantly reduced f
L occupational radiological exposures, and reduces the - potential for restraint to thermal        expansion- due to malfunctioning snubbdrs.                                                                                I 1
  -  Energy dissipation due to materia,1 and structural damping in a piping system responding to an earthquake is usually approximated                        '
in dynamic analysis by specifying an equivalent amount of viscous                    -
damping. Due to .a lack of- understanding of the parameters                          (
affecting damping, lower bound values have been mandated for use                        l in seismic design as identified in NUREG-1061-(Reference 2).                            !
l l    The-  damping    values    specified  in  Regulatory            Guide  1.61          1 (Reference 3) have been used for viscous damping for all modes                        '
considered in either elastic response- spectra or time-history                          ,
,    dynamic. seismic analyses of . Seismic Category I structures or                        )
l    components. Damping values higher than those identified                in Reference (3) are allowed only if documented test data                      is provided to support the higher values.
l The Pressure Vessel Research Committee (PVRC)                recommended an interim position on damping values that are dependent on piping modal frequency. The American Society of Mechanical Engineers (ASME) incorporated the PVRC damping position (Code Case N-411) in Section III of the ASME Boiler and Pressure Vessel Code
!      (Reference 4).
GSI 119.2 corresponds to the PRC regulatory recommendation A-2 in NUREG-1061 to modify seismic damping values used in seismic design.
Amendment F l:                                    A-52                        December 15, 1989
 
CESSAR !!n%.m.
ACCEPTANCE CRITERION The acceptance criterion for the resolution of GSI 119.2, " Piping Damping values" is that the piping system dynamic analysis shall be in accordance with those recommendations of the PRC which have been implemented by the NRC, as follows:
: 1. For        response                spectrum analyses,                                                    Regulatory    Guide' 1.84, Revision 24                        (Reference                5),                        endorses the use of piping damping values from ASME Code Case N-411 as an alternative to using values from Regulatory Guide 1.61.                                                                            Damping values in an analysis must be consistent from one of these two sources, a mixture from both is not acceptable.
: 2. The endorsement of ASME Code Case N-411 damping values for use in response spectrum analyses is subject to limiting conditions enumerated in Regulatory Guide 1.84.                                                                                        g l
: 3. For time history analyses, Regulatory Guide 1.84 does not                                                                              f endorse the use. of the code case values, and hence                                                                                      ,
Regulatory Guide 1.61 damping values shoulqi be used.
RESOLUTION                                                                                                                                      ,
The System 80+ reactor coolant system (RCS) main loop piping is                                                                                }
analyzed using time history methods. Equipment and piping damping                                                                              7 values used for this analysis are in accordance with Regulatory                                                                                a Guide 1.61.
Design and analysis of Sy. stem 80+ piping systems other than the RCS main loop use either time history or response spectrum analyses procedures as appropriate.                                                                            When time history methods are used, damping values are in accordance with Regulatory Guide 1.61. When response spectrum methods are used, Regulatory Guide 1.61 values- or the frequency dependent damping values specified in Code Case N-411 are used. When Code case N-411 damping values are used, they are used completely and consistently for the piping          system            being            analyzed,                                    and              the  relevant    limiting conditions in Regulatory Guide 1.84 are complied with.
Since the damping values used for piping design and analysis are l                  in accordance with current regulatory guidance, this issue is L                  resolved for the System 80+ Standard Design.
REFERENCE
: 1. NUREG-0933,                  "A Status Report on Unresolved Safety Issues,"
U.S. Nuclear Regulatory Commission, April 1989.
Amendment F A-53                                                      December 15, 1989 i
  ~ e . ,- , - - - -      . . _ -      , . . _ . .,          , , _ .          _ . _ _ _ . , _ . _ , _ _ _ _ . , _ _ _ , _ , _ ,
 
CESSAR1!nincarion E
: 2.        NUREG-1061, Volume 5, "R& port of the U.S. Nuclear Regulatory                                                    l Commission Piping Review Committee," U.S. Nuclear Regulatory Commission,' April 1985.
: 3.        U.S. Nuclear Regulatory Commission Regulatory Guide 1.61,
              " Damping Values for Seismic Design of Nuclear Power Plants,"
October 1973.
!  4..      American Society of' Mechanical Engineers, Boiler &-Pressure                                                    '!
Vessel Code,: Section III                        (Nuclear), American Socieby of Mechanical Engineers.
: 5.        U.S. Nucleer Regulatory Commission Regulatory Guide                              1.84,                      i
              " Design and~ Fabrication Code Case Acceptability",                                  ASME Section III, Division 1," Revision 24, June 1986.
l Amendment F A-54                    December 15, 1989
 
CESSAR nubio.
119.3:                  DECOUPLING THE OBE FROM THE BSE                                                                            j l
ISSUE                                                                                                                                                          I Generic Safety Issue (GSI) 119.3 in NUREG-0933 (Reference 1),
addresses tha question of assuring the public safety during seismic events in a more rational manner by eliminating the requirement to relate the Operating Basis Earthquake (OBE) to the magnitude of the Safe Shutdown Earthquake (SSE).
Appendix A to 10 CFR Part 100 (Reference 2) requires that nuclear power plants be designed to both the OBE and the SSE. The SSE is defined 7.4 that earthquake which produces the max $ mum vibratory ground              motion    for                which  certain        structures,                                    systems,    and components are designed to remain functional to assure:                                                                        (1) the integrity of the reactor coolant pressure boundary, (2) the                                                                                                  -
capability to shut down the reactor and maintain it in a safe condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in unacceptable offsite exposure.
The CBE is defined in Section III (d) of Appendix A as that earthquake which produces the vibratory ground motion ' for which l'    those featuras of the nuclear power plant necessary for continued                                                                                            ,
,    operation without undue risk to the health and safety of the l    public are designed to remain functional.                                                            Section V(a) (2) in Appendix A to 10 CFR Part 100 states that the OBE shall be specified by the applicant after considering the seismology and l-    geology of the region surrounding the site. Further, it states that the OBE shall be at least one-half the magnitude of the SSE.
The level of the OBE is, therefore, directly coupled with that of the SSE.
Current regulations were developed assuming that the SSE would control the design in nearly all aspects and that the OBE Would                                                                                                l serve as a separate check on those systems where continued operation was desired at a lower level of ground motien.                                                                              In                      I addition,                seismic design for                            OBE          accounts                              for  certain                        ,
safety-related factors such as                                        fatigue and seismic anchor                                                              I movement that are. not considered in the design for the SSE.                                                                                                  J However, in practice, defining the OBE as one-half the SSE l    together with assumed load factors, damping considerations, stress levels, and source limits has caused the OBE, rather than the SSE, to control the design for many systems.
GSI 119.3, which corresponds to NRC Piping Review Committee (PRC) regulatory recommendation A-3 (Reference 3), addresses decoupling
,    of the OBE from the SSE on the basis that:
l Amendment F A-55                                                        December 15, 1989 l
 
              -6is.i -          .r. m          +a_  m  _:                    --.2---- 9 .s.4          _a  s,  .
LCESSAR !!n% mon                                                                                                                                !
: 1. There is no. technical reason for -coupling the OBE with the .                                                                            I SSE,                                                                                                                                      i
: 2. Designing systems to the SSE is sufficient to ensure safety,                                                                              '
: 3. The OBE provides additional margin by specifying the' level                                                                              ,
at which inspections.are required befcre continued operation is permitted, and
: 4. Decoupling of the OBE levels and frequencies from those of the SSE will allow assurance of public safety to be placed on a more rational basis.
ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI 119.3 are that the level of OBE be defined in accordance with 10 CFR 100 Appendix A, Section III(d), and that the SSE be defined in                                                                          "
accordance with Section III(c).                                                                                                                  i However, consistent with regulatory recommendation A-3 of the NRC                                                                                1 PRC for decoupling the OBE from the SSE, the magnitude of the OBE                                                                                I need not be constrained-to at least one-half the magnitude of the SSE as required by 10.CFR 100, Appendix A, Section V(a) (2) .                                              ,
RESOLUTION System 80+ is designed for an SSE peak ground acceleration of 0.30g and an encompassing set of soil profiles which envelope the                                                                              d majority of potential nuclear power plant sites in the United States.        An OBE peak ground acceleration of 0.10g has been i
adopted for the System 80+ Standard Design and is based on a best                                                                                ;
estimate recurrence interval of 200 - 500 years, sufficient to                                                                                  ;
protect the utility investment by ensuring continued plant                                                                                      '
operation.      Establishment of the OBE and SSE is such that the SSE controls the seismic design for the majority of System 80+ plant                                                                                j structures',      systems, and components.                              The seismic design' of                                              'l saf ety ' structures, systems and components is in accordance with                                                                              I
!      Standard Review Plan Sections 3.7.1, 3.7.2, 3.7.3 and 3.7.4                                                                                      )
(Reference 4).
The System 80+ design                          implements the NRC Piping Review Committee's recommendation to decouple the OBE from the SSE; this decoupling will be reviewed by the NRC staff as part of the
!      review of CESSAR-DC. Any deviations from the requirement of 10
: j.      CFR 100 Appendix A, Section V(a) (2) will be authorized through                                                                                  i
!-      the  System 30+,            Design Certification rulemaking or other                                                                            l appropriate means. GSI 119.3 is, therefore, resolved for the l      System 80+ Standard Design.                                                                                                                      ;
'.                                                                                                                                                      l l
Amendment F L                                                        A-56                                      December 15, 1989 l
 
CESSAR ina"ic no.
E REFERENCES
: 1. NUREG-0933, " A Status Report on Unresolved Safety. Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 7. Code of Federal Regulations, Title 10, Part 100, Appendix A,
            " Seismic and Geologic Siting Criteria for Nuclear Power                                        ,
4          Plants", Office of the Federal Register National Archives                                      l and Records        Service General Services Administration,                                    ;
January 1, 1979.                                                                                l
: 3. NUREG-1061,      " Report            ef    the U.S. Nuclear          Regulatory Commission Piping Review Committee,'"U.S. Nuclear Regulatory Commission, (Volume 5) April 1985.
: 4. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear' Regulatory Commission.                                                        ]  ,
l T'
b v
Amendment F A-57              December 15, 1989
      .                              . ~ . , . . .                        - . . . - . . - . . -    .--
 
f i
i g
CESSAR !!nLm.
l' l
122.21          INITIATING FEED-AND-BLEED ISSUE Generic Safety Issue (GSI) 122.2 in NUREG-0933 (Reference 1)                                                                                          y addresses the Loss of All Feedwater Event with. respect to the                                                                                        l improved
,  provision              of    enhanced =                operator                            training  and l  instrumentation to aid the operator in determining that the plant has experienced a total loss of feedwater.
I  During routine operation at the Davis-Basse nuclear power generating. station, a loss of all' feedwater event occurred.
Subsequent to the loss of feedwater, the operators delayed initiating              feed-and-bleed                          to                    cool      the  core          on      the presumption that auxiliary feedwater flow was imminent.
p  An analysis of this event revealed that- in addition to the operators' hesitancy to commence feed-and-bleed operations, the normal control room instrumentation was found to be inadequate to L  alert the operators that feed-and-bleed was required.
The safety concern with GSI 122.2 is that a loss of all feedwater l
coupled with a failure to diagnose and take corrective action i  immediately (i.e. establish feed-and-bleed), could result in a loss of' core cooling and, therefore, jeopardize the health and safety of the public.                                                                                                                                  .
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 122.2 is that t  there shall be instrumentation and displays of sufficient l  quality, range, and reliability to enable the plant operators to
(  recognize quickly a Loss-of-All-Feedwater event and to assess when to initiate mitigating measures such as feed-and-bleed. In
    ' addition, emergency procedures guidelines should be provided to aid the operators in diagnosing the event in order to accomplish a safe _ plant shutdown.
RESOLUTION The System 80+ Standard Design incorporates the Nuplex 80+
l  Advanced Control Complex (ACC) which includes the Post-Accident l  Monitoring Instrumentation (PAMI).                                                        The PAMI is designed in accordance with the intent of the guidance given in Regulatory Guide 1.97, Rev. 3 (Reference 2)                                                              This instrumentation is itemized in CESSAR-DC, Section 7.5.1.1.5 and Table 7.5-3 and includes the parameters monitored, the number of sensed channels, sensor            ranges,    indicated                        range,                    location,    and          equipment qualification requirements.
Amendment F A-58                                    December 15, 1989
 
CESSAR1!n bi.                                                                                                                        ,
Examples        of plant parameters . monitored that are needed to identify a Loss-of-Feedwater event are: steam generator pressure and level (wide range); main and emergency feedwater flow; and reactor coolant pressure, temperature and degree of subcooling.
The feed-and-bleed function for beyond-design-basis events is performed by the use of the Safety Depressurization System (SDS) in conjunction with the                      Safsty Injection System (SIS)                                          as described in CESSAR-DC, Section 6.7. Tne PAMI also monitors and displays        SDS  and  SIS    parameters                  following                            initiation      of feed-and-bleed..
The Nuplex 80+ ACC, which both monitors normal operating and accident conditions is' designed to display the plant status to the operators in a clear and concise form.                                                                                          <
The System 80+ Standard Design also incorporates a dedicated safety-related Emergency Feedwater System,                                                      as described. in            ;;c L      CESSAR-DC, Section 10.4.9.                    This      system                  is            not  required for              ;-
normal operation but significantly reduces the probability of a Loss-of-Feedwater event occurring.
In addition to the above design features, Combustion Engineering assists the. utility owner-operator by providing . emergency procedure guidelines, as contained in report CEN-152 (Reference 3),    for the preparation of procedures. This report is further                                                            ,
discussed in the response to GSI I.C.1.
Since (1) the Nuplex 80+ Advanced Control Complex incorporates adequate and reliable instrumentation for the rapid detection of a  Loss-of-Feedwater        event          by  the        plant                      operators        and    for monitoring the subsequent cetions to achieve a safe shutdown, and (2), emergency procedurs guidelines are provided for this event, this safety issue is resolved for the System 80+ Standard Design.
REFERENCEE
: 1.      NUREG-0933,    "A Status Report on Unresolved Safety Issues",
o              U.S. Nuclear Regulatory Commission, April 1989.
: 2.      Regulatory Guide      1.97,          Revision            3,              " Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant and Environs During and Following an Accident", May 1983.
: 3.      CEN-152,    Revision 3, " Emergency                            Procedure                    Guidelines",
Combustion Engineering, Inc.
Amendment F A-59                                                December 15, 1989
  . - .    . . _ . __        __      ~ , _ _        _          _ _ _.._ _ _ _ _.. - - - _ . .                        __        __
 
t CESSARnn%.ns 124t  AUXILIARY FEEDWATER SYSTEM RELIABILITY ISSUE Generic Safety Issue        (GSI). 124    in NUREG-0933        (Reference      1),
addresses Emergency Feedwater System reliability and-availability.
* and its impact on mitigating core-melt frequency.                                      *
',            For existing plant        designs,    the  function    of    the  Emergency-        ,
.'            Feedwater (EFW) system is to supply water to the secondary side                    '
L            of the . steam generators (SG's) during various plant evolutions I            including, system fill, plant heatup, hot standby and cold shutdown. Also, the. system is designed for use subsequent to such
* design.brasis events'as loss of secondary inventory, whether this l            loss is due to normal power supply failure or due to such li            postulated accidents as a feedwater line break, a steam line break, or steam generator tube rupture.
L Industry has experienced failures of the EFW system including the loss of . all feedwater event at the Davis-Besse nuclear power generating station. This event prompted an extensive review of EFW- reliability      and      availability      by  the    NRC. Operating
            . experience, together with NRC and industry studies,                  indicates' L
that the EFW systems continue to f ail at a high rate. Various I            ' studies have also damonstrated that the EFW system continues to                        4 play a crucial role in reducing the postulated core-melt frequency.
L            Therefore, to assure a high level of EFW system availability and l            -reliability, the NRC proposed a revision to SRP 10.4.9, which L              stated that the unavailability for all operating plants and future plants should be no more than 1 x E-4 per demand after accounting for: - ETW support systems, common cause failures, and operational errors.      Furthermore, the NRC proposed that this reliability goal should be - demonstrated by PRA calculations consistant with the guidance in SRP 10.4.9, Rev.            2.
ACCEPTANCE CRITERIA The acceptance . criterion for the resolution of GSI 124, is that c              the Emergency Feedwater System shall be designed so that its l-            unavailability is no more than 1 x E-4 per demand after L              accounting for: EFW support systems, common cause failures, and l              operational errors. Furthermoro, this reliability goal should be demonstrated by PRA calculations consistent with the guidance provided in SRP 10.4.9, Rev.      2.
Amendment F A-60                      December 15, 1989 l'
 
3 i                                                                                                                            ~,
I'i CESSARin#,cui ,
1 l
RESOLUTION                                                                                                              .I The System 80+ Standard Design Emergency Feedwater System (EFW).
is        designed    to maintain a high level of availability and reliability consistant with its importance as a safety system.                                                            .
The reliability and: design features are described in CESSAR-DC, Section          10.4.9,  and include two independent trains with each
* train aligned to supply'its respective steam generator.
Each train consists of:
: 1.          'one emergency feedwater storage tank (EFWST),
    -2.            one 100 percent capacity motor driven pump subtrain and one 100 percent capacity steam driven pump subtrain,
: 3.            flow control valve,
: 4.            isolation valve,
: 5. .          check valve,
: 6.            a cavitating venturi, and
                                                                                                                          -a
    =7.-          specified instrumentation.                                                                                  r one          design  feature        of            the                EFW system which  improves              its    f reliability is its component and piping separation and diversity.
For example, each subtrain is separated from the other end therefore has its own discharge line through the steam generator.                                                      -
isolation valve and check valve. In addition, the pump crossover lines contain redundant, locked - closed, isolation valves. The subtrain design reduces the potential for single failure and                                                          E improves system reliability.                                                                                                .
Because of the - improved reliability of the Emergency Feedwater System design, the unavailability for the system was estimated from.FRA studies to be in the range of 1 x E-4 to 1 x E-5 per-demand as described in CESSAR-DC, Section 10.4.9.1.2. Analysis identified : in - CESSAR-DC, Appendix 10A, - which was developed using generic data assesses the systems ability to function . on demand                                                          -
and' demonstrates its compliance with the above unavailability range.            Therefore,      the                EFW ; system meets the recommended unavailability goal of 1 x E-4 per demand identified in SRP Section 10.4.9, Rev. 2.
Since the Emergency Feedwater System meets the recommended unavailability goal specified in SRP Section 10.4.9, Rev.2, Subsection II, paragraph Sc, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1.          NUREG-0933,    "A Status Report on Unresolved Safety Issues",
U. S. Nuclear Regulatory Commission, April 1989.                                                            >
Amendment F A-61                  December 15, 1989
 
CESSAR !annemor                                                                                    .
125.I.03t -SAFETY PARAMETER DISPLAY SYSTEM AVAILABILITY l
l ISSUE Generic Safety Issue (GSI) 125.I.03 in NUREG-0933 (peference 1),
addresses safety Parameter Display System (SPDS) availability and L the reliability of the information it displays.
The TMI-2 accident demonstrated the need for improving how information is relayed to the control room operators. As a result,        NUREG-0737,      I.D.2,        (Reference  2)    required        .the
  . installation of a SPDS.          The      purpose of the SPDS is to ' improve how information is provided to the control room operators by l supplying them with continuous information from which the plant safety status can be readily and reliably assessed.
However, after installation of the SPDS at operating plants, the Davis-Besse plant Loss-of-Feedwater event and other operating plant SPDS availability surveys raised concerns regarding SPDS reliability and availability and its impact on plant safety.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 125.I.03 is that the availability of the SPDS should be such that it can accomplish its intended function as described in NUREG-0737, I.D.2,      (i.e.,      to provide the control room operators with a                        ;
continuous means of determining the plant safety status).
RESQLUTION In the System 80+ Standard Design, SPDS functions are performed by the Advanced Control Complex (ACC).                        The ACC uses an integrated          information display hierarchy to present both safety-related and non-safety-related plant data for use by the control        room operators        (See        CESSAR-DC,  Section    7.5). An                  e integrated system ensures that the operator will be familiar with information displays during abnormal transients,                        since the operator uses the same displays for both normal and abnormal operations.
In the ACC, SPDS functions are implemented by three distinct information display systems regularly used by the operatcr: the Integrated Plant            Status    Overview (IPS0) panel, the Data Processing System (DPS), and the Discrete Indication and Alarm System (DIAS).            These display systems have been designed and configured (as described in CESSAR-DC Sections 7.7.1.4,                      7.7.1.5, and 7.7.1.7) such that the loss of any one of them does not result in a total loss of necessary information.
Amendment F A-62                December 15, 1989
 
l l
  ^    CESSAR !!n*nemo,,                                                                                    j L
l l
l The IPSO ' panel receives data from both the DIAS and DPS via-                                      I l_
different          data    links.      The IPSO keeps operations personnel                        )
        . informed        about        the status of the plant's critical                  safety          '
functions and success paths as described in CESSAR-DC Section 18.7.1.2.            It    also provides          a  limited set of key plant parameters.            Implementation of the IPSO panel hardware considers redundancy for enhanced reliability.
The DPS is configured redundantly for improved reliability.                            It acquires plant data (e.g., process variable and component status)                                  ,
validates it, and executes applications programs for its display                                    i page hierarchy.                The portion which addresses SPDS requirements                      '
l        includes:            IPSO, critical safety functions, and success path l        monitoring            to      aid    the    operator    in    gathering      supporting
!        information and- problem diagnosis (see CESSAR-DC, ' Section 18.7.1.8.2). This is the primary means of implementing the SPDS functions in the ACC.                                                                            i w
Figures          7.7-16        and 7.7-17        in  CESSAR-DC    show    the    basic  w i configuration of the DIAS design.                        The DIAS employs discrete            ;
indicators that are used to display validated safety and                                    .
non-safety-related                plant    process      parameters    including    those required by the SPDS functions. It uses a segmented design -to provide a degree of hardware independence and fault resistance between various segments. The DIAS channel P (DIAS-P) segment is                              ,
designed .to            be      physically    separate    from  and      electrically      :
independent of the remaining DIAS channel N (DIAS-N) segment and                            i the DPS such that a single failure will not cause a. loss of more-                            :
than one of the three display methods (DIAS-P, DIAS-N or DPS).
In summary, the-SPDS functions identified in NUREG-0737, I.D.2, are performed by IPSO, DPS, and DIAS in the Advanced Control Complex. Each system incorporates improved design features such as separate and redundant hardware, power supplies (including battery backup), and system self-test features. These design features assure that the IPSO, DIAS, and DPS are very reliable thus ' minimizing the availability concern associated with the SPDS. Therefore, this issue is resolved for the System                                80+
Standard Design.
REFERENCES
: 1.      NUREG-0933,          "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.      NUREG-0737, " Clarification Of TMI Action Plan Requirements",                            l U.S. Nuclear Regulatory Commission.                                                      f
: 3.      Generic Letter No. 82-33, Supplement 1 to NUREG-0737,                      U.S.          l Nuclear Regulatory Commission.
Amendment F A-63                    December 15, 1989          l 1
1
 
CESSARUmnemo.
128: ELECTRICAL POWER RELIABILITY ISSUE Generic Safety Issue        (GSI)      128 in NUREG-0933    (Reference        1),    .
addresses      the    reliability        of    onsite  electrical        systems.
NUREG-0933 combined three GSI's previously individually listed under . NUREG-0737 (Reference 2) in order to provide a .more                            ,
integrated approach to resolving these interrelated issues.
ACCEPTANCE CRITERIA The acceptance' criteria        for the          resolution of GSI        128  are encompassed in other GSI's namely - 48, 49, and A-30,                    which are given in NUREG-0933.
RESOLUTION The resolucion for GSI 128 is identified in- the responses to GSI's 48, 49, and A-30 which are addressed and resolved in CESSAR-DC, Appendix A. Since GSI 128 is subsumed by these three GSI's, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933, " A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulacery. Commission, October 1980.
l I
f l-Amendment F A-64                  December 15, 1989 i
 
CESSAR !!nincum 130t ESSENTIAL SERVICE WATER PUMP FAILURES AT                        ,
MULTI-PLANT..8ITES                              ,
ISSUE Generic Safety Issue (GSI) 130 in NUREG-0933 (Reference 1),
addresses the concern that potential core damage could occur                    '
because of insufficient cooling water flow from the station service water system (SSWS) for safety-related systems and
      . components due to shared SSWS's at multi-plant sites.
Design of the SSWS (or service water system (SWS) as identified in NUREG-0933) varies considerably among existing plants. At some multi-unit sites, portions of the SSWS are shared among the units. Multi-plant configurations for the SSWS may result in the inability to provide needed cooling water to                safety-related systems due to the unavailability of SSWS components which can be shared between units. Should the SSWS fail to provide adequate                  ,
cooling capability to shutdown a plant, when subject to a loss of            '
SSWS,-a core damage accident could result.
A related safety issue, USI 051, also requires separation and independence of SSWSs at multi-plant sites.
l ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 130, is that the SSWS shall not be shared between units of a multi-unit site.
Specifically, each unit shall be provided with a dedicated SSWS, and shall be designed to the same requirements as a single unit.
RESOLUTION The System 80+ Standard Design is a single independent plant design; that is, all systems and components necessary for the operation of the plant.are dedicated to that particular plant.
Therefore, SSWS is designed for a single unit and does            not rely In addition, on other systems or components from other unit (s) .
the SSWS.is an open cycle system consisting of 2 redundant trains l
        -(4 SSWS pumps) and is, therefore, very reliable (see CESSAR-DC, Section 9. 2.1) .        The SSWS has the capability to dissipate the heat loads necessary for a safe reactor shutdown by rejecting                    :
heat delivered from the safety-related component cooling water                  !
The  ccWS  cools    safety-related  components, system    (ccWS).
i        including those required for safe shutdown of the reactor, i
Where construction of multiple plants is desirable, . separation                j and independence of all systems and components including the SSWS is maintained by the owner-operator and the architect-engineer (see CESSAR-DC, Section 1.2.1.3).
Amendment F A-65                  December 15, 1989
 
i CESSAR !!nbri.
L
  . .                                                                                            (
i The possibility of potential core damage from a SSWS system failure.as a result of shared systems and components is minimized because of the required separation and independence both in the plant design and in the SSWS design. Therefore, this issue is.
resolved for the System 80+ Standard Design.                                                l REFERENCES 1._  NUREG-0933,  "A' Status Report on Unresolved Safety Issues",-
U.S. Nuclear Regulatory Commission, April 1989.
t
                                      '                                                            j Amendment F A-66                              December 15, 1989
 
CESSAR n!Nicari I
1-02t    ABYMMETRIC BLOWDOWN LOADB ON REACTOR PRIMARY COOLANT SYSTEMS                                                              I ZKE9R Unresolved Safety Issue (USI) ' A-02 in NUREG-0933- (Reference 1),                                              :
addresses asymmetric blowdown loads imposed on the reactor vessel (RV) as a result of a design basis loss of coolant accident (LOCA). The resultant forces from these loads could affect a      - reactor vessel support integrity, thus jeopardizing plant safety.
I        A break in a large reactor coolant pipe could cause several rapidly occurring internal and external transient loads to act                                                  1 upon the reactor vessel. In the event of a postulated LOCA at the vessel nozzle, asymmetric I.cCA loading could result from forces                                              i' induced on the reactor internals by transient differential pressures across'the core barrel and by forces on'the vessel due to  transient differential pressures in the reactor cavity.
Differential pressures, although of short duration, could place significant loads on the reactor vessel supports,                                        thereby affecting their integrity.
The    NRC    reviewed    these  predicted  asymmetric            loadings                    and developed acceptance criteria and guidelines, which have been documented in NUREG-0609,- (Reference 2).          In 1987, a " broad scope" revision of General Design Criterion 4 allowed the Leak-Before-Break (LBB) methodology documented by the NRC in NUREG-1061, Volume 3 (Reference 3), to be applied to all high energy piping.
ACCEPTANCE CRITERIA The    acceptance      criterion  for the resolution of USI                                  A-02
        '(documented in NUREG-0609)          is that the design of the reactor primary coolant system shall demonstrate that the asymmetric loads on the reactor vessel, internals, primary coolant loop, and components shall not exceed the limits imposed by the applicable codes and standards.
RESOLUTION The 1987 " broad scope" revision of General Design Criterion 4 permits application of the LBB methodology for all high energy pipes.in nuclear power plants. The LBB methodology of NUREG-1061, Volume    3,  is - endorsed by the NRC in implementing the " broad scope"    rule.      The application of this methodology in the evaluation of the System 80+ Standard Design reactor coolant piping is descrit-4 in CESSAR-DC, Section 3. 6.3. (see also, GSI 119.1)
Amendment F A-67                    December 15, 1989
                                                              . . ~ . . _ _ _ _ . . . - - , - _ .
 
i t
CESSAR E hno,.
Where    the            LBB    approach      cannot    be          applied                    effectively,      a determination of pipe break locations and dynamic effects is made.      These are identified in CESSAR-DC, Section 3.6.2. The
* criteria used to define pipe break and/or crack , locations and configurations              are      given      in    CESSAR-DC,                            Section    3.6.2.1.
Postulated ruptures are classified as circumferential breaks, longitudinal breaks, leakage cracks, or through wall cracks. Each postulated rupture                  is    considered separately as a single                                                            ,
postulated initiating event.
The Leak-Before-Break methodology in CESSAR-DC, Section                                                    3.6.3, was used to demonstrate that the detection of flaws in pipes can                                                                        '
be- assured before they cause large break LOCAs and, therefore, l the asymmetric loads and the resultant loads on primary system components and supports are no longer significant. Any effects from    small-break                LOCAs      which    cannot                            be    eliminated    by Leak-Before-Break methodology, are accounted for in the faulted condition analysis by applying a conservative factor to the safe-shutdown-earthquake resultant loads.
In summarn all loads are within limits imposed by industry codes and. standards for the primary coolant system and, therefore, this issue is resolved for the System 80+ Standard Design.                                                                                -
REFERENCES
: 1. . NUREG-0933,            "A Status Report on Unresolved Safety Issues",
U. S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG-0609, " Asymmetric Blowdown Loads on PWR Primary Piping Systems", U.S. Nuclear Regulatory Commission, November 1980.
: 3. NUREG-1061, Volume                  3,    " Evaluation of Potential for Pipe Breaks", U.S. Nuclear Regulatory Commission, November 1984.
Amendment F A-68                                            December 15, 1989
_    .. _ . _ . . - .      _      __  -.        .      , _ ___._.. _.~._.. - .._._.                      _ _ . _ . _ . _ _ . _
 
CESSAR tilhu.
1-09t kNTICIPATED TRANSIENTS WITHOUT SCRAM, (ATWS) 181 3 Generic Safety Issue (GSI) A-09 in NUREG-0933                  (Reference 1),
addresses the issue of assuring that the reactor can attain safe shutdown after incurring an anticipated transient with a failure of the Reactor Trip System (RTS).                    An ATWS is an expected operational transient (such as a loss of feedwater, loss of condenser vacuum, or loss of offsite power to the reactor) which is accompanied by a failure of the RTS to shut down the reactor.
LcenPTANCE CRITEU&
The acceptance criterion for the resolution of GSI A-09 is that the reactor must be capable of reaching a safe shutdown condition as identified in 10 CFR 50.62 (Reference 2), after incurring an                  t
,        anticipated transient and a RTS failure.
Specifically                            -
1    'v
                . meet section (c) (1) ("the mitigation requirement") of 10 CFR 50.62,              plant equipment uglst automatically initiate
                ' emergency feedwater and turbine trip under conditions indicative of an ATWS.                  This equipment must function reliably and must be diverse and independent from the RTS.
: 2. To meet section (c)(2) ("the prevention requirement") of 10 CFR 50.62, the plant must have a scram system which is diverse and independent from the existing RTS.
RESOLUTION The System 80+ Standard Design contains safety and control grade systems designed to protect the plant and mitigate the consequences of design basis events. These systems have the following design features:
: 1. The Plant Protection System (PPS) consists of the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) (see CESSAR-DC Section 7.1.1.1).
The PPS is designed with both redundancy and diversity to maximize the ability to mitigate transients.                  However, should an ATWS occur, the System 80+ standard Design includes    an          Alternate    Protection System (APS)      for mitigation.
i Amendment F A-69              December 15, 1989
 
f I
e i
CESSAR tiiWiem.                                                                                                                                                                              !
: 2.          The APS augments the RPS to address                                                                                            10          CFR  50.62                          l requirements for the reduction in risk of ATWS and for the                                                                                                                        ;
use of ATWS Mitigating Systems Actuation Circuitry (ASMAC) .                                                                                                                      '
t The APS design includes an Alternate Reactor Trip Signal                                                                                                                          *
(ARTS) and Alternate Feedwater Actuation Signal. (AFAS) that are separate and diverse from the PPS (CESSAR-DC, Section                                                                                                                        I
                                                                                                                                                                                                            ~
L                          7. 7.1.1.11) .                            The APS equipment provides diverse and independent mechanisms to reduce the possibility of an ATWS and to provide additional assurance that an ATWS event could                                                                                                                      l l                          be mitigated.
: 3.          The ARTS will initiate a reactor trip when pressuriger l                          pressure exceeds a predetermined value (see CESSAR-DC, Table l                          7.7.1). Turbine trip signals can also initiate ARTS if the Reactor Power Cutback System is out of service.                                                                                              The ARTS
!                          turbine trip input is manually enabled from the main control l                          panel.
l l                          The ARTS circuitry is diverse and independent from that of
                  .      the RPS.                The ARTS design uses a two-out-of-two logic to open the motor-generator output contactors, thus removing motive power to the Reactor Trip Switchgear System (see                                                                                                                            ,
CESSAR-DC, Figure 7.7-12).                                      In addition, ARTS provides for a                                                                              ,]
turbine trip through a relay which is not part of the PPS or                                                                                                                        ;
APS.                                                                                                                                                                                j
: 4.          The AFAS will start emergency feedwater to a sh nm generator                                                                                                                        '
when the level in that steam generator decreases below a L                          predeter.ained value (see CESSAR-DC, Table 7. 7.1) . Its L                          circuitry is diverse from that of the RPS.                                                                                Actuation of the l                          EFW (pumps and valves) is achieved by sending isolated AFAS l
signals to the Engineered Safety Feature Component Control System described in CESSAR-DC, Section 7.3.
In summary, the System 80+ Standard Design includes a control-                                                                                                                                  !
grade APS that supplements the RPS and provides a diverse and                                                                                                                                  j independent means of reactor trip.                                                          Also, the APS supplies a                                                                          j control grada AFAS which maintains a diverse and independent method of automatically initiating emergency feedwater.                                                                                                      Since the APS is designed to meet 10 CFR 50.62 as identified in CESSAR-DC Section 7.7.1.1.11, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1.        NUREG-0933,                  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulutory Commission, April 1989.
Amendment F A-70                                                                    December 15, 1989 l
1
    --e .-4,-      r _,.  ,.,-_..,e.
                                            . , - ,    ..,,.,,._.-.m.      ,    ,,,,m,        _ . . , , , , . . . , , , , , , . . . , , , , , , , , _ , , _ _
 
o CESSAR imificui..
: 2. Code of Federal Regulations, Title 10, Part 50, Section 62,
    " Requirements  for  Reduction  of  Risk    From Anticipated Transients    Without    Scram      (ATWS)      Events    for Light-Water-cooled Nuclear Power Plants",        Office of the Federal    Register    National    Ar: hives    and    Records Administration, June 26, 1984.
A Amendment F A-71                  December 15, 1989
 
c                                                                              y CESSAR !!nkn.S 1-12:    FRACTURE TOUGENESS OF STEAM GENERATOR kND REACTOR COOY M PUMP SUPPORTS                            l ISSUE Unresolved Safety Issue (USI) A-12 in NUREG-0933 (Reference 1),
addresses minimizing the susceptibility for lamellar tearing and                l low fracture toughness of major reactor coolant system (RCS)                    j component supports.
During the course of licensing the North Anna Units 1 and 2,              a number of questions were raised as to the potential for lamellar              j tearing      (a  cracking    phenomenon    that  occurs  beneath  welds    '
involving rolled steel plate) and low fracture toughness of the                l steam generator and reactor coolant pump (RCP) support materials.              .
Concerns regarding the supports at North Anna have been found to              ,
apply to all PWR's.
With regard to lamellar tearing, the results of an extensive                  .
literature      survey conducted by Sandia, see NUREG/CR-3009 (Reference 2) and discussed by the NRC in NUREG-0577 (Reference
: 3)    concluded    that,    although lamellar tearing        is  a  common occurrence in        structural    steel  construction,  virtually no.
documentation exists describing in-service failures of nuclear power plant supports from lamellar tearing. The NRC also stated in NUREG-0577        that preliminary research conducted by EPRI concluded that lamellar tearing is generally detected and corrected during construction and that a reasonable safety factor on strength can bound experimental results governing the lamellar tearing phenomenon.        Subsection NF of the ASME Code therefore,          '
provides for adequate toughness of reactor coolant system supports.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of USI A-12 is that the major RCS component supports must meet the requirements f    specified in Subsection NF of the ASME Code (Reference 4).
RESOLUTION System    80+    Standard    Design  Reactor  Coolant  Pump  and  Steam Generator supports are designed in accordance with 10 CFR 50.55a (Reference 5) which further references accepted industry codes, including the ASME Code.
The relevant ASTM material            specifications are    identified  in CESSAR-DC, Table 5.2-2.
Amendment F A-72                  December 15, 1989
 
CESSAR !Me.no.
System 60+ Standard Design steam generator supports consist of a sliding base bolted to an integrally attached conical skirt which is mated to the staatn generator (see CESSAR-DC, Section 5.4.14.2, Paragraph B).      The steam generator also has lateral supports which    are identified in CESSAR-DC, Figure 5.4.14-3 (upper supports). The reactor coolant pump supports are provided with four vertical support columns, four horizontal sup aort columns, and two horirontal snubbers (see CESSAR-DC, Sect,on 5.4.14.2, Paragraph C).
Major component supports for the reactor coolant system are designed and fabricated in accordance with the ASME Code, Section III, Subsection NF, as described in CESSAR-DC, Section 5. 4.14.
, Thus,    " code" materials are used in the fabrication of the supports; consequently the fracture toughness of these materials is    in accordance with code requirements.                This issue        is, therefore, resolved for the System 80+ Standard Design.
REFEREMCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG/CR-3009,      " Fracture    Toughness    of    PWR    Component Supports",  U.S. Nuclear    Regulatory    Commission,        February      a 1983.
i l 3. NUREG-0577;    Revision      1,    " Potential for Low Fracture                -
I        Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports",          U. S. Nuclear Regulatory Commission, October 1983.
: 4. American    Socisty      of  Mechanical      Engineers,      Boiler      &
Pressure Vessel        Code,    Section  III    (Nuclear),      American Society of Mechanical Engineers.
: 5. Code of Federal Regulations, Title 10, part 50, Section 55a,  " Codes  and    Standards",    Office    of    the    Federai Register National        Archives    and  Records      Administration, January 1, 1988.
Amendment F A-13                    December 15, 1989
 
I r
CESSAR !!uhn.
t A-13t    SNUBBER.0PERABILITY ASSURANCE IB8UE Unresolved Safety Issue (USI) A-13 in NUREG-0933 (Reference 1),                                  .
addresses snubber selection and operability for safety related systems and components by identifying the need for:                                              ,
a consistent means of determining                  snubber                              >
: 1)                                                                  operability through standardized functional testing;
: 2)    a set of criteria for selection and specification; and, l
: 3)    preservice and inservice inspection programs.
Snubbers    are    utilized        primarily  as seismic and pipe whip                        4 restraints at operating plants.                Their safety function is to l  operate as rigid supports for restraining the motion of systems or components under dynamic load conditione such as earthquakes and severe hydraulic transients, e.g., pipe breaks.
According to NUREG-0933, a substantial number of Licensee Event Reports (LER's), concerning snubber operability, were issued by utilities.      A review of these LER's showed that a variety of methods were employed to determine the operability of the snubbers and that different types of snubbers were used for l  systems with similar configurations.
l ACCEPTAMCE. CRITERIA l
The acceptance criterion for the resolution of USI A-13 is that l  the    design,        specification,        installation,    and    in-service l
operability of snubbers must meet the intent of the guidance l  given in SRP Section 3.9.3 (Reference 2).
Specifically, during the design of safety systems or components for which snubbers are to be used,                sufficient consideration should be given as to their unique application, i.e., their response to normal, upset, and faulted conditions and the effect of these responses on the associated system and/or component.
RESOLUTION For the System 80+ Standard Design, snubbers are minimized by using design optimization procedures                (see CESSAR-DC,          Section 3.9.3.4. However, where required, snubber supports are used as shock arrestors for safety-related systems and components.
Snubbers are used as structural supports during a dynamic event Amendment F A-74                  December 15, 1989
 
l                      CESSAR !!hno,.
i                        such as earthquake or pipe break, but during normal operation act
-                        as  passive            devices which                    accommodate                    normal  expansions    and contractions without resistance.
Assurance of snubber operWility fer the System 00+ Standard Design    is                provided by incorporating analytical, design, i                installation, in-service, and verification criteria. The elements of snubber operability assurance includet
: 1. Consideration of load cycles and travel that each snubber
=                              will experience during normal plant operating conditions.
: 2. Verification that the thermal growth rates of the system do not exceed the required lock-up velocity of the snubber.
E L                        3. Appropriate                    characterization                                of    snubber    mechanical
;                              properties                    in            the        structural                    analysis    of    the snubber-supported system.
l
: 4. For engineered, large bore snubbers, issuance of a design specification to the snubber supplier,                                                      describing    the
;                              required                structural            and              mechanical          performance    of  the                            -
snubber; with subsequent verification that the specified design and fabrication requirements were met.
: 5. Verification                  that          snubbers                      are    properly  installed    and operable prior to plant operation, through visual inspection and      through                        measurement                      of    thermal    movements    of snubber-supported systems during start-up tests.
m                        6. A snubber in-service inspection and testing program, which o                              includes                  periodic            maintenance and visual                            inspection, y                              inspection following a faulted event, a functional testing
-                              program,                and repair or replacement of                                    snubbers    failing r
inspection or test criteria.
In summary, during the design of safety-related systems or components for which snubbers are to be used, sufficient consideration is given as to their unique application, (i.e.,
their response to normal, upset and faulted conditions snd the effect of these responses on the associated                                                        system and/or component). Thus the design, specification, installation, and in-service operability of snubbers meets the intent of SRP Section 3.9.3 and this issue is resolved for the System 80+
_                      Standard Design.
REFERENCES r
: 1. NUREG-0933,                "A Status Report on Unresolved Safety Issues",
Z                            U.S. Nuclear Regulatory Commission, April 1989.
Amendment F A-75                                  December 15, 1989
      =
 
4 CESSAR !BM,en...
I I
: 2. NUREG-0800, " Standard Review Plan for the Review of Safety                                            <
Analysis Reports for Nuclear Power Plants -- LWR Edition",                                            i U.S. Nuclear Regulatory Commission.                                                                    ;
i l 1 J
1 1
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                                                                                                                    )
I I
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Amendment F                              1 A-76                    December 15, 1989 l
 
CESSAR M en.                                                                                                                                    l t
1-25                    McNa8AFETY LOADS ON CLASS 1E POWER SOURCES I
issue i
Gs.neric Safety Issue (GSI) A-25 in NUREG-0933 (Reference 1),
addresses the potential safety degradation of a Class 1E power                                                                                  ,
system caused by its connection te a non-safety-related power                                                                                    :
source or load.                                                                                                                                  .
t There are two approaches to assuring the reliability of the                                                                                      '
safety-related system Class 1E power supplies for future plants.
The first approach is to restrict the connection of primarily safety loads to Class 1E power supplies. (In previous designs, non-safety electrical equipment we.s connected to Class 1E power supplies      (i.e.,                        the emergency diesel generators) to provide a source of power during Loss-Of-of fsite-Power (LOOP) events.)
The  second                      approach                      is      to    limit    the    connection                  of non-safety-related electrical equipment to the Class 1E power systems and assure that when this equipment is connected to the                                                                                  -
Class 1E power system ' that the equipment and the connections conform      to              the                    requirements          for  independence,              electrical                        ;
These requirements are isolation,              and' physical separation.
identified in IEEE Standard 384-1981 (Reference 2), and guidance                                                                                '
is provided in Regulatory Guide 1.75, Revision 2 (Reference 3).
[ Supplemental information on Class 1E safety systems may be found in IEEE Standard 603-1980, ANSI N42.7-1972 and IEEE St1tndard 308-1980, (References 4, 5 and 6 respectively).]                                                                                                  ;
Both industry and the NRC, through IEEE Standard 384-1981 and                                                                                    )
Regulatory              Guide                          1.75,      have      determined    that  these                  design requirements provide an acceptable means of achieving an adequate                                                                                (
level of reliability for the class 1E power supplies.                                                          Therefore, a commensurate level of safety for the safety systems is assured.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI A-25 is that the reliability and level of safety of Class 1E power sources and the safety systems which they supply may not be degraded by the sharing          of                    loads                  between      safety-related      systems                  and non-safety-related systems.
l Specifically,                      the second approach,                          identified in the issue statement, shall be used in establishing an acceptable level of reliability                  and                  safety        for      Class    1E  power sources and safety-related systems.
Amendment F A-77                      December 15, 1989
        --      .... - .-..-.-. - .-.- . ~ . _ _ . . . . . .                                            . - ..~ - -.. .          . - - . . . .  -
 
l, CESSARnnLam.                                                                          )
This shall be accomplished by assuring that the interface between safety-related and non-safety-related equipment on Class 1E power sources and safety-related systems is adequately controlled by                        l meeting the independence, electrical isolation, and physical separation requirements identified in IEEE Standard 384-1981 and other applicable        standards,      References      2    and  4  through  6, respectively, taking into consideration the guidance provided in Regulatory Guide 1.75, Revision 2.
RESOLUTION The System 80+ Standard Design incorporates the second approach for assuring the reliability and adequate level of safety for the Class  1E power sources            and safety-related systems by the selective connection of non-safety-related equipment and strict control of the interface between the non-safety-related equipment and Class 1E power system.
The    System    80+      Standard      Design      contains    safety-related instrumentation and controls and supporting systems which are essential
* for the safe operation and shutdown of the reactor.
These systems are identified in CESSAR-DC, Section 7.1.1. Each                        i safety-related system conforms to the requirements of IEEE Standard 384-1981 and General Design Criteria 3 and 24, and meets the intent of Regulatory Guide 1.75.                  Design requirements have been specified such that the power supply and each safety system t
have the same reliability.            These requirements are described in CESSAR-DC, Sections 8.1.3 and 8.1.4 and refer to IEEE Standard 384-1981 as supplemented by Regulatory Guide 1.75 Rev. 2.                            ',
Section      8.3.1.2.7      also      addresses      IEEE    Standard 279-1971 (Reference 7) system level requirements for safety related Class 1E power supplies and equipment.                                                      ,
In the System 80+ Standard Design, a separate emergency power supply provides power to non-safety-related equipment during
  " LOOP" events. This separate power supply, which is designated the    Alternate      AC    Source,          minimizes    the    connection    of non-safety-related equipment to Class 1E power supplies.                        The power source reliability is thereby improved.
For the limited cases in which non-safety-related equipment must be connected to Class 1E power supplies, the System 80+ Standard Design adheres to the special design requirements which have been adopted by the NRC.          These requirements are identified in IEEE Standard 384-1981 and promulgated in Regulatory Guide 1.75 and in the Standard Review Plan (Reference 8).                    The bases for these requirements are physical separation, electrical isolation, and circuit independence.
Since both the safety systems and their Class 1E power supplies conform to      the  requirements            of  IEEE  Standard    384-1981  and Amendment F A-78                      December 15, 1989 t
 
i f
CESSAR mr'ificari =
l meet the intent of Regulatory Guide 1.75, Rev.                              2,  an acceptable level of safety exists for both the safety systems and their                                                  !
Class 1E power supplies. Therefore, this issue is resolved for                                                i the System Go+ Standard Design.
BEFEREMCES
: 1. NUREG-0933,    "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. IEEE Standard 384-1981, " Criteria for Separation of Class 1E Equipment and Circuits", The Institute of Electrical and Electronics Engineers, Inc.
: 3. Regulatory Guide 1.75, Rev. 2, " Physical Independence of Electric Systems", U.S.                            Nuclear Regulatory Commission,                        '
September 1978.
: 4. IEEE  Standard      603-1980,                    " Standard  Criteria for Safety                  -
Systems    for  Nuclear              Power            Generating      Stations",    The          :
Institute of Electrical and Electronics Engineers, Inc.
: 5. ANSI 'N42.7-1972,              " Criteria              for Protection        Systems for Nuclear Power Generating                            Stations",
American National              ,
Standards Institute.                                                                                  .
: 6. IEEE Standard 308-1980, " Criteria for Class 1E Electric                                            -
Systems    for  Nuclear              Power Generating Stations",                    The              .
Institute of Electrical and Electronic Engineers, Inc.
: 7. IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations,"                                The Institute        of Electrical and Electronic Engineers, Inc.
: 8. NUREG-0800, " Standard Review Plan for The Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
I l                                                                                                                      l
!                                                                                                                      I Amendment F A-79                            December 15, 1989
 
i CESSAR inlineamn                                                                                !
1-261  #Ir_1CTOR YESSEL PRESSURE TRANSIENT PROTECTION ISSUE                                                                                            !
Unresolved Safety Issue (USI) A-26 in NUREG-0371 (Reference 1),
deals with ensuring that adequate protection from Reactor Vessel                                .
Pressure Transients is provided.
According to NUREG-0933 and NUREG a0244 (References 2 and 3) there                              :
have been,    since 1972,    over 30 reported events of pressure transients which have exceeded the pressure-temperature limits of                                :
pressurized water reactor vessels. These limits are identified in the technical specifications for each vessel and are based on the requirements of Appendix G to 10 CFR 50 (Reference 4).                            The majority of these events occurred at relatively low reactor vessel temperatures at which the material has less toughness and is  more    susceptible    to  failure    through    brittle                fracture.
Therefore, the margin of safety to vessel failure under low temperature conditions is reduced. The safety margin will be further reduced by the reduction in the toughness properties of
    . the vessel caused by neutron irradiation.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of USI A-26 is that                                  ,
adequate measures for protection of the reactor vessel from pressure transients shall be included in the system design.
Specifically,      the  acceptance      criteria    for  the        overpressure protection system are based on meeting the intent of the relevant
: l. guidance identified in SRP Section 5.2.2 Rev. 2 (Reference 5).
i l      Specific acceptance criteria necessary to meet the requirements l      of General Design criteria 15 and 31 in 10 CFR 50 Appendix A (Reference 6) are as follows:
A. For overpressure protection during power operation of the reactor, the relief valves shall be designed with sufficient capacity to preclude actuation of safety valves                              during l
l normal operational transients.
Safety valves shall be designed with sufficient capacity to limit the pressure to less than 110% of the reactor coolant pressure boundary design pressure (as specified by the ASME B&PV Code in Reference 7), during the most severe abnormal operational transient and with the reactor scrammed.
Also, sufficient margin shall be available to account for uncertainties in the design and operation of the plant.
Amendment F
!                                            A-80                    December 15, 1989 i
 
e f
1 CESSAR Wario,,                                                                                                                          .
                                                                                                                                            /
l 5
I B. The low temperature, overpressure protection (LTOP) system shall be designed in accordance with the requirements of Branch Technical Position RSB 5-2 (Reference 5) . The LTOP                                                                          ,
system shall be operable during startup and shutdown conditions below the enable temperature.
RESOLtrfION The System 80+ Standard Design incorporates a variety of methods to assure the integrity of the Reactor Coolant System (RCS). The RCS is designed in accordance with 10 CFR 50.55a (Reference 8) which further references the ASME B&PV Code and other accepted industry codes and standards.                    In addition, the reactor coolant                                                        ,
l  pressure boundary is defined in accordance with ANSI /ANS 51.1 (Reference 9).                                                                                                                          .
Furthermore, there are a number of systems and components that                                                                    '
ensure the integrity of the RCS, namely:                                                                                            ,
: 1. Reactor Protective System                                                                                                  .A Initiates    a          reactor      trip    to            protect the RCS pressure boundary in the event of high pressurizar pressure, see CESSAR-DC, Section 7.2.
: 2. Reactor Coolant System Components a)      Primary Safety Valves                                                                                                        j These are sized using a conservative Design Basis                                                                            l Event, namely a loss of turbine load with a delayed reactor trip and in accordance with the ASME B&PV code to limit the pressure boundary design pressure; see CESSAR-DC, Section 5.2.2, Appendix 5A.
b)      Large Volume Pressurizar The RCS pressurizer has a larger volume than the System                                                                      '
l 80  pressurizer which permits a                                                        larger range          of transients and reduces the challenges to safety valves; see CESSAR-DC, Table 5.4.10-1.
c)    Steam Generator Secondary Safety Valves These valves are conservatively sized to pass excess steam flow to limit steam generator pressure to less than 110% of steam generator design pressure during the worst case transients; see CESSAR-DC, Section 5.2.2,                                                                          l Appendix 5A (submitted June 1990).
Amendment F A-81                                                          December 15, 1989
 
CESSAR in#ican..
1 l
: 3. Shutdown' Cooling System (SCS)                                                                                                  l l
This system provides for overpressure protection of the RCS                                                                      j at        reduced          temperatures                    and pressures                              and    therefore addresses the LTOP issue. The over-pressure protection for                                                                      '
the RCS is accomplished by providing a relief path through I
special- LTOP relief valves incorporated in the SCS for the RCS during heatup and cooldown.                                            The LTOP relief valves are                          I sized and adjusted to the appropriate setpoint(s) to ensure adequate overpressure protection for the RCS at reduced                                                                        :
temperatures and pressures; see CESSAR-DC, Section 5.2.2.10.
                                                                                                                                                    +
System 80+ Standard Design in designed to provide adequate                                                                          i overpressure protection for the RCS/ Reactor Coolant Pressure                                                                      l Boundary by incorporating the systems and components described above. This issue is, therefore, resolved for the System 80+
Standard Design. NRC closeout will be automatic upon successful completion of the review of CESSAR-DC Sections 5.2.2 and Appendix 5A.                                                                                                                                l REFERENCES
: 1. NUREG-0371,              " Task    Actior Plans for Generic Activities (Category A)",, U.S.              Nuclear Regulatory Commission, November 1978.
: 2. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 3. NUREG-0244 Rev.                1, " Technical Report on Reactor Vessel Pressure Transients", U.S. Nuclear Regulatory Commission,                                                                      ,
May 1978.
: 4. Code of Federal Regulations, Title 10, Part 50, Appendix G,
                    " Fracture Toughness Requirements",                                                      Office of the Federal Register National                    Archives and Records Administration,                                                      t January 1, 1988.
.              5. NUREG-0800, " Standard Review Plan for the Review Of Safety Analysis Reports for Nuclear Power Plants-- LWR Edition",
U.S. Nuclear Regulatory Commission.
: 6. 10 CFR 50 Appendix A, " General Design Criteria", Office of the            Federal      Register,              National                            Archives        and  Records Administration.
: 7. ASME Boiler and Pressure Vessel Code, Section III, Article NM-7000, " Protection Against Overpressure", American Society of Mechanical Engineers.
Amendment F l                                                                    A-82                                                  December 15, 1989
  . _ . _ . _ _          __ _ . _ . ~ , _            .._          ._ _ _ . _ _ _ _ . , _ _ _ ___ _ _ _ , _ . , _ _ _ . . _
 
CESSAR !HWien..
: 8. Codo of Federal Regulations, Title 10, Part 50, Section 55a,
      " Codes and Standards", Office of the Federal Register National Archives and Records Administration, January                                  1, 1988.
: 9. ANSI /ANS 51.1, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", 1983.
:u i
Amendment F A-83 December 15, 1989
~
 
t r
CESSAR iniincui..                                                                                                    .
i 4
A-29        NUCu1_R POWER Pn_WT DESIGN FOR TE2 REDUCTION OF YULNERABILITY TO INDUSTRIAL 8ABOTAGE l
l ISSUE Generic Safety Issue (GSI) A-29 in NUREG-0933 (Reference 1),                                                        I addresses the susceptibility of nuclear power plants to                                                            ,
industrial sabotage, the resulting risk to plant safety, and the countermeasures to assure an acceptable level of protection.                                                        ,
consideration should be given to sabotage during the design phase of the plant. The goal would be to achievt an acceptable level of protection of a plant to induscrial sabotage by emphasizing                                                        .
design features which reduce the likelihood of the plant I
4 incurring damage from industrial sabotage, both internal and                                                        i external.
New design features (e.g., relocating emergency feedwater tanks                                                    '
to protected areas, increasing the monitoring, separation and independence of plant protection systems, providing additional back-up sources of power) which pitovide countermeasures to                                                    ,
sabotage must be consistent with plant safety requirements.
j          ACCEPTANCE CRITXRIA The acceptance criterion for the resolution of GSI A-29, is that plants shall be designed to be resistant to the effects of                                                          ,
internal and external sabotage through prevention, deterrence, and mitigation.
Specifically,            plant      safety-related              systems          and        components            >
1          required for the safe operation and shutdown of the plant shall be designed for protection against and mitigation of sabotage.
RESOLU'1 ION l          The System 80+ Standard Design is configured to be sabotage resistant (See CESSAR-DC Chapter 13, Appendix 13A).
This is accomplished in various ways including:
l
: 1. locating safety-related equipment                          in        secure        areas  and            ,
controlling personnel access;
: 2. designing for separation, independence, and redundancy of safe shutdown and support systems; 3.. monitoring          of      equipment              status        continually            (e.g.,
automated-testing of equipment) ;
Amendment F A-84                        December 15, 1989 l
 
i t
CESSAR 2%.m l
: 4. evaluation and improvement of                        safe shutdown systems        to                      ,
provide for              damage control            measures in the event          of sabotage; J. plant layout design to accommodate a variety of access soning schemes; and                                                                                        ,
: 9. protection of design features located outside of plant buildings and structures which are more vulnerable to                                                      ,
sabotage.
s Included in the System 80+ standard Design are a number of impreved design features which enhance plant resistance to                                                      :
sabotage compared to those of a more traditional design:                                                        ,
                                                                                                                      ?
      -    Redesign of the emergency feedwater system (E N) to function                                                !
in the event of a loss of offsite and onsite power. This design accommodates coincident failure of a single active mechanical or electrical component or the offacts of a high                                              '
l          or moderate-energy pipe rupture.
      -    EW storage tanks (2) which are designed so that each has a capacity to support full flow for 30 minutes, if the                                                        ,
initiating event is a main feedwater line break. The water                                                  ,
supply available is capable of maintaining the plant at hot standby for eight hours and than provide for an orderly cooldown to shutdown cooling entry conditions.                                                          }
      -    Locating the emergency feedwater storage tanks within the                                                  '
auxiliary building where access is restricted.
l
      -    Addressing station blackout by designing the steam-driven EW components such that they are capable of providing EW to the steam generators coincident with a single failure.
Battery backed power is used to assure steam-driven pump discharge valves are open. The. battery also powers the                                                    _
turbine governor speed control and steam generator level                                                    ;
indication to provide control of steam generator level.
l Also, an alternate AC power source                          is provided as an additional power source.
      -    Providing the shutdown cooling system with containment spray                                                -
system piping cross-connects that permit the containment spray pumps to supplement the shutdown cooling system should a shutdown cooling system pump become                            inoperative.      In addition, each containment spray and shutdown cooling system pump is physically separated from the other.
I Amendment F A-85                                December 15, 1989
 
i l
CESSAR !!nincu.                                                                                                    l
        -    Increasing the design presaurs and piping schedule of the shutdown cooling system to 900 psia so that even if subjected to full RCS pressure, the pressure boundary does                                                  -
not fail.
        -    Addition of the Safety Depressurization System (SDS) which                                                  '
permits an alternate and physically reparate decay heat I
removal path (i.e. once through cooling) in the event the preferred decay heat removal system is' disabled.                                                            ;
        -    Relocating        the refueling water storage tank into                                          the        L containment thus restricting access and reducing                                                  the      ,
likelihood of sabotage.
l
        -    Placing safety-related equipment (e.g., emergency safeguards                                                -
components) within the containment sub-sphere, where access                                                t is strictly controlled.
        -    Limiting access to the control room. The Nuplex 80+ Control                                                ;
complex is designed to limit access to the control room. The                                                ,
design reduces the number of required personnel needed in the controlling workspace, yet provides direct monitoring                                                  !
(visual and via the data processing CRT's) of the operater's actions and state of the plant from adjacent supervisory l-            offices.
        -    Designing the Nuplex 80+ instrumentation and controls to incorporate semi-automated and on-line testing features for the Plant Protection System and on-line monitoring of fluid and electrical systems, thus enhancing the detection of sabotage.
        -    Enhancing detection of plant sabotage through improved plant l              system monitoring. The Nuplex 80+ Data Processing Syctem l              Design includes computer aided testing and success path monitoring of emergency safeguards features systens which i
aid        in  detecting              abnormal conditions                    and/or system degradation which may result from attempted sabotage.                                                      t
        -    Designing the Nuplex 80+ instrumentation and controls to provide      channel              separation                for many      systems    including separate equipment rooms for each safety channol. With adequate access control to each channel,                                            this design increases the difficulty in equipment sabotage.
In summary, the System 80+ Standard Design is highly resistant to sabotage because of the design features described, which protect against both internal and external sabotage.                                        Therefore,        this issue is resolved for the System 80+ Standard Design.
Amendment F A-86                                December 15, 1989
    - --    - - -  . --        - - - - - . - .          .~ . --. - _..- -                                        .
 
e                              -                  -
    'l  0.);~
g                                                                                      l j r,- >- .    -
3 iM ,[l                                                                                                        l l
CESSAR tilA.                                                                                      l l
I REFEREMCZE J
: 1.        WUREG-0933,  "A Status Report on Unratsolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
I j
l
                                                                                                                )
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4 l
:n            ,
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                                                                                                                )
J L            ,
I i
Amendment F A-87              December 15, 1989
 
CESSAR Mu%an.
1-30t    ADEOUACY OF BAFETY-RELATED                                ,
DC POWER SUPPLIES IMUE                                                                                            ,
Unresolved Safety Issue (USI) A-30 in NUREG-0933 (Reference 1) ,
addresses the reliability of DC power supplies used in the control and actuation of safety-related components and systems.                                  8 DC power systems in nuclear power plants provide control and motive power to a variety of safety-related equipment including valves, instrumentation, emergency diesel generators, and many other components and systems. This power is needed during abnormal shutdowns and accident situations, as well as during normal operations. Presently, a minimum of two divisions of DC power are . required to supply control and motive power to this                                  ,
safety-related equipment; failure of one division would generally cause    a  reactor              scram for this type            of          configuration.
Furthermore,              if  the independence of the            two        divisions is compromised through the failure of a bus-tie breaker to function properly, then a fault in one division could propagate to the redundant division resulting in a loss of redundancy.
The safety significance of a loss of the two divisions is that a total loss of DC power supplied to safety-related equipment could occur thus prohibiting the equipment from performing its intended
! safety function.                                                                                ,
ACCEPTANCE CRITERIA l
The acceptance criterion for the resolution of USI A-30 is that the reliability of safety-related Dc power supply systems shall be improved.
Specifically, the present configuration of DC power supplies used
!  in current plant designs should be replaced by a more reliable configuration on new plants using the following criteria: (1) All non-safety related loads should be placed on completely separate non-safety-related DC power systems, and (2) the safety-related (class 1E) DC supplies shall be divided into four physically and electrically separated systems satisfying the requirement of General Design Criterion 17 in 10 CFR 50 Appendix A (Reference 2) l for independence to reduce the probability of reactor trip in the event of the less of a single safety-related DC bus.
RESOLUTION The System 80+ Standard Design provides dedicated DC electrical buses for non-safety-related electrical loads and meets criterion l
l 1 listed above, as described in CESSAR-DC, Section 8.3.2.
Amendment F A-88                            December 15, 1989
 
r CESSAR inMncamn                                                                                                i l
The requirement for reducing the probability of reactor trip in the event of a loss of a single safety-related bus (criterion 2) is met by the System 80+ Standard Design, also as described in detail in CESSAR-DC, Section 8.3.2. As shown in CESSAR-DC, Figure                                              ,
8.3.2-2,    the safety-related (Class 1E) DC power supply system consists of four separate isolated channels. The DC bus from each channel can be isolated from its battery bank and alternately supplied from its division's DC bus. For typical Class 1E DC and                                              >
AC instrumentation and control power supply systems, either the Channel A or Channel C DC bus can be supplied from the Division I, also a 1E DC bus. Similarly, either the Channel B or Channel D bus can be supplied from the Division II, also a 1E DC bus.                                                    .
Cross-ties between buses, however, are isolated through two sets                                              ;
of manually operated fusible disconnects.
Furthermore,      during      normal operation,          both        switches          are maintained    in  the    open  position. This    method          of        isolation                  '
satisfies the independence criterion of 10 CFR 50, Appendix A                                            a (General Design Criterion 17). Thus, in the safety-related (Class                                          .
1E) DC electrical distribution system design,                  the two safety                                  ,
divisions are both ehysically and electrically isolated from each other and from the non-safety-related (non-class 1E) division.
That is, for this electrical distribution configuration, there are      no    bus      cross-ties      between        safety-related                  and-                  i non-safety-related buses, and therefore, a fault in one division cannot propagate to the redundant division.                                                                ,
d In    summary, the safety-related (Class 1E) AC and DC bus configuration      meets      the  criterion        for    non-safety                and safety-related      load    separation.      This      bus        configuration, therefore, reduces the likelihood of a reactor trip (in the event of a loss of a single safety-related (Class 1E) DC bus), and thus meets the requirements of 10 CFR 50,                  Appendix A (GDC 17).
Therefore, this issue is resolved for the System 80+ Standard                                                    1 Design.                                                                                                          l REFERENCES
: 1. NUREG-0933,    "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear Power Plants", Office of The Federal Register, National Archives and Records Administration.
l l
l Amendment F A-89                      December 15, 1989
 
CESSAR !!nhm.
l 1-311      RESIDURL nay_ REMOVAL (RRR) SHUTDOWN REOUIREMENTS IBRUE i
Unresolved Safety Issue (USI) A-31 in NUREG-0933 (Reference 1),
addresses the safe shutdown of the reactor, following an accident or abnormal condition other than a Loss of Coolant Accident (ICCA) , from a hot standby condition (i.e., the primary system is at or near normal operating temperature and pressure) to a cold shutdown condition. Considerable emphasis has been placed on long-term cooling which is typically achieved by the residual heat removal system which starts to operate when the reactor                                                    ,
coolant pressure and temperature are substantially lower than the hot-standby values.
I Even though it may generally be considered safe to maintain a reactor in a hot-standby condition for a long time, experience has shown that there have been abnormal occurrences that required long-term cooling until the reactor coolant system was cold enough to perform inspection and repairs.                              For this reason, the ability to transfer heat from the react'or to the environment, after    a      shutdown resulting from an accident or abnormal occurrence, is an important safety function.                                It is essential that a power plant be                        able to go              from hot-standby to cold-shutdown conditions subsequent to any accident or abnormal occurrence condition.
ACCETTANCE CRITERIA The acceptance criterion for the resolution of USI A-31 is that the RNR system shall be designed so that the reactor can be brought from a " Hot Standby" to a " Cold Shutdown" condition as described in SRP Section 5.4.7 Rev. 3 (Reference 2).
Specifically, the RHR system shall meet the intent of                                    the following functional requirements with respect to cooldown:
: 1. The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems.              These systems shall satisfy 10 CFR 50 Appendix A (Reference 3) General Design Criteria (GDC)                                1 through 5.
: 2. The system (s) shall have suitable redundancy in components and      features,    and                suitable        interconnections,      leak connection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available)                            the  system function can be accomplished assuming a single failure.
Amendment F A-90                      December 15, 1989
 
l CESSAR tinificui .
: 3. The system shall be capable of being opert ted from the control room with either onsite or offsite power available.
In demonstrating that the system can perform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable,          ". f suitably justified.
: 4. The system (s) shall be capable of bringing the reactor to a cold shutdown condition, with either offsite or onsite power available, within a reasonable period of time following a shutdown, assuming the most limiting single failure.
In addition to the functional requirements listed above, there are certain additional requirements for the RHR system including, pressure relief, pump protection, test and operation.
Also, there is a requirement that the emergency feedwater system shall have sufficient inventory to permit operation at het                -
shutdown for at least 4 hours, followed by a cooldown to the              M conditions permitting operation of the RHR system. The inventory needed for cooldown, shall be based on the longest cooldown                -
required with either onsite or offsite power available with an assumed single failure.
RESOLUTION
                                                                                ?
The System 80+ Standard Design utilizes the Shutdown Cooling System (SCS), the Reactor Coolant Gas Vent System (RCGV), the Safety Depressurization System (SDs), the Atmospheric Dump Valves (ADV), and the Emergency Feedwater Systems (EFW) as the preferred means to bring the reactor plant from hot standby to a cold shutdown condition within s reasonable period of time. These safety-related systems are normally operated from the control room and are described in CESSAR-DC, Sections 5.4.7,      10.1, and 10.4.9.
To reduce the reactor coolant system (RCS) temperature from hot standby to cold shutdown requires two phases.      The initial phase of cooldown is accomplished by heat rejection from the steam generators (SG) to the atmosphere using the ADV.
Plant pressure is normally reduced by use of the Auxiliary Pressurizar Spray System (non-safety-related). If this system is not available, the pressurizar can be depressurized by venting the pressurizer to the Drain Tank or the Incontainment Refueling Water Tank via the RCGV or SDS, respectively.
Amendment F A-91                  December 15, 1989.
 
l t
CESSAR iH#icama                                                                                        j t
After the RCS conditions have been reduced to approximately 350 degrees F and 550 psia, the SCS is put into operation to further reduce and maintain RCS temperature at the refueling temperature (see CESSAR-DC, Section 5.4.7.1). The SCS and the SG atmospheric                                        f steam release and EFW are used to cooldown the reactor following                                        .
a IOCA.                                                                                                ;
Non-LOCA accidents, such as stsam or feedwater line breaks, and steam generator tube ruptures also are accommodated by the SCS                                          -
(see CESSAR-DC, Section 5.4.7.1).                                                                        i j
All of these systems are designed to meet GDC 1 through 5 for
;      quality assurance, protection against natural phenomena,                                fire protection,    environmental and      missiles,              and shared            systems,            ;
structures and components (see CESSAR-DC, Sections 3.1.1, and                                          i 5.4.7.1.3 paragraphs P, D, Q, F, and M).
l The SCS is highly reliable and is designed to perform its safety                                          '
function assuming a single failure by incorporating suitable                                            f redundancy in components and power (see CESSAR-DC, Sections 5.4.7.1.2, 5.4.7.1.3 and 5.4.7.2.5).                    For example, the SCS is designed with redundant components and power supplies such that the RCS can be brought to refueling temperature (cold shutdown)                                          !
using one of two redundant SC,S trains.                                                                  1 The EFW system is also highly reliable and is designed to function in the event of a loss of offsite and onsite power. The                                        )
design accommodates a coincident failure of a single active                                              j mechanical or electrical component.            In addition, the EFW system                              !
is capable of maintaining the plant in a hot standby condition                                          j for eight hours and then providing for an orderly cooldown to shutdown conditions (see CESSAR-DC, Section 10.4.9.1.2).
In addition to normal offsite power sources,                      physically and electrically independent and redundant emergency power supply systems are provided to power safety related components (see                                              l CESSAR-DC, Sections 8.1.2 and 8.2).                                                                      l The shutdown cooling system is designed to meet the intent of of SRP Section 5.4.7, Rev. 3 with respect to providing a means to                                          i bring the reactor plant from het standby to cold shutdown under all accident or abnormal occurrence conditions, as described above.      Therefore, this issue is resolved for the System 80+                                          l Standard Design.
REFERENCES
: 1.      NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
Amendment F A-92                            December 15, 1989
 
                                                                                  \
                                                                                  \
CESSAR WWicui.
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: 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
* U.S. Nuclear Regulatory Commission.
: 3. 10 CTR 50 Appendix A, " General Design Criteria for Nuclear Power Plants",  office of The Tederal Register,        National Archives and Records Administration.                                    ;
                                                                                  }
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l Amendment F A-93    December 15, 1989
 
I CESSAR !!i#ican.
1-361      CONTROL OF HEAVY LOADS NEAR SPENT FUEL l
ISSUE Unresolved Safety Issue (USI) A-36 in NUREG-0933 (Reference 1),
addresses the consequences of dropping heavy loads,                                    (i.e.,      loads that weigh more than the combined weight of a fuel assembly and its handling tool) on spent fuel, including the potential for fuel damage and/or criticality which could result in radiological releases.              Also, USI A-36 addresses the possibility of damage to safety-related equipment which could prevent safe shutdown of the reactor or jeopardize core decay heat removal.                                                                                      .
ACCEPTAMCE CRITERIA                                                                                                                l l      The overall acceptance criterion for the resolution of USI A-36                                                                    :
                                                                                                                                          ^
l      is that the nuclear plant design shall consider the ef fects of the movement of heavy loads over spent fuel and equipment important to achieving and/or maintaining safe shutdown of the reactor, consistent with the guidance of the Standard Review Plan 9.1.5 (Reference 2).
l' Specifically, the load handling systems are acceptable if (1) safe load paths are defined through procedures and operator training so, that to the extent practical, heavy loads are not l
carried over or near irradiated fuel or safe shutdown equipment;
,      (2)      adequate        consideration              is  given to operator training, l      handling            component design,                  load    handling instructions,                  and equipment inspection to insure reliable operation of the handling system;          (3) mechanical stops or electrical interlocks are incorporated to prevent movement of heavy leads over irradiated fuel or in proximity to equipment associated with redundant core shutdown paths; and (4) the load handling system conforms to the relevant requirements of General Design Criteria (GDC) 2, 4, 5, l      and 61 and References 3-7.
RESOLUTION The System 80+ Standard Design addresses the above criteria as follows:
: 1.        The component (Heavy Load) handling procedure guidelines will require the owner-operator to establish the safe load path and perform special handling component inspections prior to lift.
: 2.        The      plant      operating              procedure      guidelines    will      require appropriate operator training and crane inspections.
Amendment F i                                                              A-94                    December 15, 1989
 
i CESSAR !imne.n..                                                                                                                        i i
: 3.      The cask handling crane is provided with mechanical stops                                                                      3 and electrical interlocks to prevent its movement near the                                                                      '
spent fuel pool after the pool contains irradiated fuel (See CESSAR-DC, Section 9.1.4).                                                                                                      ;
: 4.      The new fuel handling crane is provided with mechanical                                                                        '
stops and electrical interlocks to restrict its motion between the new fuel shipping container receipt area, the new fuel inspection and storage areas, and the new fuel elavator (See CESSAR-DC, Section 9.1.4).                                                                                        ,
: 5.      The spent fuel building is arranged so that the spent fuel
  <                    cask does not pass over critical components during its passage from the shipping vehicle to the cask laydown area (see CESSAR-DC, Sections 9.1.4.1.3 and 9.1.4.3.1).
: 6.      The reactor vessel head lift rig and the reactor vessel internals lift rigs are designed in accordance with the                                                                    m    .
acceptable (stress) factors of safety (Reference 4).                                                                      c    )
: 7.      An analysis of a drop of the reactor vessel head onto the                                                                  c    ,
reactor vessel .is performed as described in CESSAR-DC Section 9.1.4.3.3,                                and the results are shown to be acceptable.                                                                                  ,
I            8.      An analysis of the upper guide structure drop on the reactor L                      vessel is performed to demonstrate that this event                                                    is                  :
bounded by the result of the analysis of Item (7) above.                                                                  +
: 9.      The load handling system is designed in accordance with the relevant requirements of GDC 2, 4, 5, and 61 (See CESSAR-DC,                                                                    >
Stetion 3.1 and 9.1.4) and the guidance of References 3-7.
Since the acceptance criteria for this safety issue are met, the issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. -    NUREG-0933,          "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.      NUREG-0800, " Standard Review Plan for The Review Of Safety Analysis Reports for Nuclear Power Plants",                                  U.            S. Nuclear Regulatory Commission.
: 3.      ANSI-N14.6, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 KG) or More for Nuclear Materials".                                                                                              ;
1 l
l Amendment F A-95                  December 15, 1989 l
1
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CESSAR !!Micario.                                                        )
: 4. NUREG-0612,      " Control of Heavy Loads at      Nuclear  Power Plants", U.S. Nuclear Regulatory Commission.
: 5. ANSI /ANS-57.2, " Design Requirements for Light Water Reactor        i spent Fuel Storage Facilities at Nuclear Power Plants".
: 6. ANSI /ANS-57.1, " Design Requirements for Light Water Reactor Fuel Handling Systems".
: 7. NUREG-0554, " Single Failure Proof Cranes for Nuclear Power            i Plants", U.S. Nuclear Regulatory Commission.
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l-Amendment F A-96              December 15, 1989 l                                                                                1
  .-.        . . - . . ,      - . .          -  w-,~. -        .    ,
 
LCESSAR Memor 1-43t  CONTAINMENT EMERGENCY SUMP PERFORMANCE XABB Unresolved- Safety- Issue (USI) A-43 in NUREG-0933 (Reference 1),
                                    . addresses the technical concerns raised when LOCA generated debris is introduced into the containment - sump, and the impact this debris _ and the potential for air ingestion have on the
                .                      containment spray system and components.
The' Emergency safeguards Systens (including Safety Injection and containment Spray) are actuated following a I4CA and, depending upon' the severity of the event, the Containment Spray System (CSS)- and the Safety. Injection System (SIS) may be-configured to support long ters. cooling of the containment and/or reactor core.
Long.- term cooling _ is accomplished by recirculation of the spray water recovered from the containment building by various building drains which direct water into the containment sump.                      The recirculated water must be sufficiently free of debris and air so that containment spray , pump performance. is not impaired and long-term      recirculation    flow    capability    is  not    seriously .
: degraded.
The specific concerns are:
: 1.      Containment    sump hydraulic performance under post-LOCA adver.se  conditions    resulting    in  a    potential    vortex formati t,- air- ingestion and subsequent failure of the recirc O cion pumps.
: 2. The    transport    of  large    quantities    of  LOCA-generated debris to the sump screen (s) and potential for blockage, reducing net positive suction head (NPSH) margin below that required for the recirculation pumps to maintain long-term cooling.
: 3. The - capability of the CSS pumps to continue pumping with fineidebris trapped in' pump seals and bearings.
Adequate recirculation cooling capacity is necessary to prevent core-melt and containment rupture following a postulated LOCA.
ACCEPTANCE CRITERIA The- acceptance      criterion    for  the  resolution    of  USI  A-43, (Reference 1) is that the containment spray and safety injection systems  (including the ' containment sump)        must be capable of
                                        - performing their intended safety functions (in accordance with the requirements of GDC 38),          including long-term core and/or containment building cooling.
Amendment F A-97                  December 15, 1989
      = _ . _ _ _ _ _ _ _ . . . . . . . . . .
 
b CESSAR !!n% n w                                                                                                                                  ,
The containment spray and safety injection systems must meet the intent of SRP Section 6.2.2, Rev. 4 " Containment Heat Removal Systems"          (Reference 2), and Regulatory Guide 1.82, Rev. 1 (Reference 3).
l  Specifically, to satisfy the requirement of GDC 38 regarding the l  long-term containment spray system (CSS)                                and safety injection system        (SIS),      the . containment sump (s)                should be designed to provide a reliable, long-term water source for the SIS and CSS recirculation pumps. Provision should be made in the containment design _to allow drainage of spray and safety injection water to the containment sump (s), and for recirculation of this water through the containment sprays and safety injection systems. The design of the sump (s) and the protective screen assemblies is a 1  critical element in assuring long-term recirculation cooling l- capability.
C Therefore, adequate design consideration is necessary for:
a)    Sump hydraulic performance" to preclude vortexing and air ingestion, b)    Evaluation of potential debris generation and associated
      ,  effects including debris screen blockage and,-
c)    RHR and CSS pump performance under postulated post-LOCA                                                                                '
conditions.
Finally, the containment sump design criteria, as identified in Regulatory Guide 1.82, Rev. 1, shall be addressed.                                            Examples of these design criteria include:
  -      Number, location, and geometry of the sump (s),
Type, size, and location of debris screens,
  -      Location of containment drains,                                                                                                          ,
  -      Containment sump effects on spray or recirculation pump NPSH, and Accessibility of containment spray system and sump (s) for inservice inspection.
RESOLUTION In the System 80+ Standard Design, Engineered Safety Features (ESF) are incorporated to mitigate design basis events (DBE's),
including a loss of coolant accident (LOCA).
Amendment F A-98                            December 15, 1989
 
CESSAR inamo,,
Two principal systems used to mitigate the effects of a LOCA are                                                      l
  -the safety injection system (SIS) (see CESSAR-DC, Section 6.3),
and the containment spray system (CSS) (see CESSAR-DC, Section 6.5). These systems utilize an In-containment Refueling Water Storage Tank - (IRWST) as their source of water, which is the equivalent- of the refueling water storage tank (RWST) and containment sump of a pre-ALWR plant.
2  The IRWST performs additional                functions beyond those of the                                            i conventional containment sump.              The.IRWST for System 80+ Standard Design provides a single source of water for both the safety injection and~ containment spray pumps.                            The IRWST is toroidal in shape and utilizes the lower section of                          the spherical containment as its outer boundary.                  The IRWST                is                  enclosed to prevent contamination and excessive containment                      humidity.
The arrangement of the IRWST within the System 80+ Standard Design containment offers advantages over conventional sumps.                                                      ,  ,
Like a sump, the IRWST is the source of water for SIS and CSS                                                          !
!  pumps, but the protection afforded the pumps against debris                                                          J ingestion or blockage is significantly greater.                                            First, water.in containment draihing back to the IRWST must pass through a vertical screen vall greater than six (6) feet high and more than forty (40) feet long into a holdup volume of several thousand cubic feet, which serves as a solids trap.                                                From the holdup volume, water overflows back into the IRWST.                                                Finally, each          ,
!  combination of SIS pump and CSS pump has its own connection to the IRWST.
Long-term return of spray water from upper level elevations is not dependent on individual floor screens and piping.                                                      Major openings, such as hatches and stairwells, are also available to return water to the screened entrance to the holdup volume.
The IRWST also meets the intent of SRP Section 6.2.2, Rev. 4 and Regulatory Guide            1.82,    Rev. I with respect IRWST hydraulic L  performance;          evaluation of potential                              debris        generation        and associated effects,                including dabris                              screen    blockage;      and preservation of NPSH for the SIS and CSS pumps after an accident l    (postulated post-LOCA conditions).              In addition, the IRWST meets the multi-sump requirement by providing multiple pathways to the IRWST for containment spray and safety injection water introduced into the containment building.                Should one drain become fouled with debris there are other drains to collect and direct water to the IRWST.
l L
Amendment F A-99                                            December 15, 1989            i
 
CESSAR unincum The IRWST has the advantage that during normal. full power operation i~c is possible to perform a full flow test of the safety injection pumps and containment spray pumps while taking suction from the IRWST and discharging back to the IRWST via a recirculation line. Satisfactory hydraulic performance of the IRWST can be verified by testing at runout conditions on the pumps and minimum level in the IRWST.
The System 80+ Standard Design Safety. Injection System meets the intent of Regulatory Guide 1.82, Rev. 1 (see CESSAR-DC, Section 6.3.1.3). The Containment Spray System        . is designed to GDC 38 (see CESSAR-DC, Section 6.5.1. 3) . The IRWST is designed to meet the intent of SRP Section 6.2.2, Rev. 4 and Regulatory Guide                ;
1.82,  Rev. 1  (see CESSAR-DC,    Section 6.3.2.2, subsection          '
6.3.2.2.1.). Therefore,  this  issue  is resolved for the System 80+ Standard Design.                                                        ,
REFERENCE 8
: 1. NUREG-0933,  "A Status Report - on Unresolved Safety Issues,"
U.S. Nuclear Regulatory Commission, April 1989.                        !
: 2. NUREG-0800, "Standar.d Review Plan for the Review of Safety
  -    Analysis Reports for Nuclear Power Plants---LWR Edition",            !
U.S. Nuclear Regulatory Commission.
i
: 3. Regulatory Guide 1.82, Rev. 1,  " Water Sources for Long-Term        l Recirculation Cooling Following a Loss-of-Coolant Accident",          !
U.S. Nuclear Regulatory Commicsion, November 1985.
e l
Amendment F A-100                    December 15, 1989 l.
 
      -CESSAR Unincanon A-45:          8HUTDOWN DECAY HEAT REMOVAL REOUIREMENTS ISSUE Unresolved Safety Issue                            (USI)  A-45          in NUREG-0933              (Reference
: 1)  addresses . the Decay Heat Removal                                    (DER)          function,      defined                  '
as the~ ability of a plant to remove residual heat from the Reactor Coolant System after a plant shutdown after normal                                                                          ,
operation              or  due- to abnormal                          events            or    Loss-of-Coolant              ,      ,
i Accidents (LOCAs),                            and  to    prepare            the          plant    for        cold shutdown conditions.
In recent years the NRC has vigorously addressed the issue of DMR improvements due to:                                first,          the Three Mile Island accident,              and  second,              the  results            of          a    DER  PRA          study performed on six operating plants in 1987.
l,
!      This' comprehensive                          PRA. DHR study concentrated on the DHR                                          ..,
systems        and- their contribution to core melt                                              frequencies.
This study assessed the consequences of                                                  both    internal          and external initiators.                          The study found that DHR-related core damage risk is in a range between 7 x E-5 and-4 x E-4 per reactor year with an' average value of 2 x E-4.                                      This, along with other results - (i.e. lack of system redundancies, lack of separation,
* general system arrangement, human errors, etc.), suggested that                                                                .
the resolution of this issue would need to be plant specific.                                                                ;
ACCEPTANCE CRITERIA l
After    the            DHR      PRA          study    was          conducted,            the    NRC        staff established a goal that core damage due to failure of the DHR function should be less than 1 x E-5 per reactor year as identified in NUREG-0933.                          This goal shall be demonstrated by a Level I Shutdown Cooling System (SCS) PRA.
RESOLUTION As part of the C-E System 80+ Standard Design certification, e      a  full Level I PRA has been conducted which included an assessment of the core damage frequency due to failure of the    DHR            (or Shutdown TheCooling                          System)            function            (see CESSAR-DC, Appendix B) .                                PRA determined the core damage frequency attributable to internal initiating events such as steam generator tube rupture and station blackout,                                                    as well as- external events such as tornadoes and                                                  earthquakes,          and was    performed to meet the requirements                                                  of    10    CFR        52 (Reference 2).
Amendment F A-101                                  December 15, 1989
 
l CESSAR E!namo, l
The    results    showed        that      the          core    damage      frequency      l for failure of        the. SCS. capability,            along    with    failure    of    I I
other systems      included        in  the      core        damage    sequences,      is lower      than  the    NRC-      requirement-            mentioned        above.      PRA assumptions are identified in- Appendix B.                                  In  addition, owner-operators      will      be    required            to    assure      that    these-    )
to -remain                                            plant      '
assumptions      continue                                valid    for      their specific reliability programs.
Since    the PRA demonstrates the ability of the System 80+
Standard Design to surpass the NRC goal for SCS-related core damage risk, contingent upon- the requirement that individual plant owner-operators meet the assumptions of the PRA to ensure                                  1
!    that the PRA remains valid during operation, this safety issue is                                )
therefore resolved for the System 80+ Standard Design.                                          l l
REFERENCES
: 1. NUREG-0933,    "A-  Status        Report            on    Unresolved      Safety Issues",    U. S. Nuclear Regulatory                  Commission,      April 1989.
                                                                                                      )
2.*    10    CFR  52,    "Early      Site          Permits;      Standard      Design      1
          , Certification;    and    Combined        Licences,        for Nuclear      Power      l Reactors",    office      of    the      Federal          Register,      National -    j Archives and Records Administration.                                                      '
                                                                                                    , l I
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s Amendment F A-102                            December 15, 1989 l
1
 
CESSAR innneuio,,
A-491  PRESSURIZED THERMAL SHOCK (PTS)
ISSUE Unresolved Safety Issue (USI) A-49 in NUREG-0933 (Reference 1) ,
addresses        reactor            vessel  integrity  under  conditions    of pressurized thermal shock.                    Reactor vessel integrity could be threatened                    by a  combination    of  neutron  embrittlement, overcooling, excessive reactor vessel pressure, and a small flaw or crack.
As plants accumulate service time, neutron irradiation reduces the material toughness of the reactor vessel.                        The fracture toughness      sensitivity                of the  vessel  material  to  neutron embrittlement is, in part, a function of the copper content of the vessel material (e.g., the higher the copper content of the base metal, the greater the sensitivity of the metal to neutron embrittlement).                  Decreased fracture toughness makes it more          4 likely that a crack already present in the vessel inner wall could grow to a size that might threaten vessel integrity, should a combination of vessel' overcooling and overpressure occur.
ACCEPTANCE CRITERION The acceptance criterion for the resolution of USI A-49 is that the requirements for pressurized thermal shock identified in 10 CFR 50.61 (Reference 2) shall be met.
Specifically, for continued reactor operation during the plants design life, without further NRC review, projected values of RT                        limiting reference temperature at the inner vessel sue $ce()the for reactor                beltline materials shall remain less than a screening value of 270 degrees F for plates, forgings, and axial weld materials, or 300 degrees F for -circumferential weld materials.
RESOLUTION The reactor vessel for the System 80+ Standard-Design is designed in accordance with the ASME Boiler and Pressure Vessel Code, Section III (Reference 3) and other accepted industry codes and standards.
To assure an adequate safety margin for future plants,                          the reactor vessel for the System 80+ Standard Design incorporates proven fabrication techniques together with well characterized steel and weld material which exhibit uniform properties and predictable behavior (see CESSAR-DC, Section 5.3).
Amendinent F A-103                  December 15, 1989
 
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i    LCESSAR !!nhm.
These material and fabrication techniques and other reactor vessel design features are described as'follows:
The ' copper content is controlled to assure that the RTPTS will remain acceptable over the life of the plant.
The characterization of the- steel and weel materials was established through . industrial and governmental ' studies which      examined the material properties in both the unirradiated and the            irradiated                  condition.          Inservice inspection and material surveillance programs are also conducted during the service life of the vessel, further ensuring adequate vessel integrity and safety margin.
j          -
Design,    materials of construction, fabrication methods, inspection        requirements,            shipment                and      installation, operating conditions, and inservice surveillance are all components of a program to assure reactor vessel integrity for the plant -design lifetime.                  A complete description of the reactor vessel design is given in CESSAR-DC, Section
                  ,5.3.
l
            -      The System 80+ Standard Design reactor vessel is fabricated from ring forgings, thus eliminating, vertical welds in the                                                      .
beltline region where neutron irradiation is greatest. The                                                        ,
i:                  elimination of these . particular welds further reduces the possibilities of impurities-in weld material which are known                                                      ,
approaches the screening i                  to result in an RT                  . that i                  criterion of 270 degra d F.
            -      Furthermore, the System 80+ Standard Design reactor vessel 4
meets the requirements of 10 CFR . 50. 61 as described- in CESSAR-DC, Section 5.2.2.11.                  Specifically, the calculated RT      at the and of the 60-year service life is 109 degrees F,Mich . is significantly below the screening criterion of 270-degrees F:for plates forgings and axial weld materials, or 300 degrees F for circumferential weld materials.
Since the System 80+ reactor vessel design complies with the ASME code and other accepted industry codes and standards, and meets the requirements of 10 CFR 50.61, this issue is resolved for the System 80+ Standard Design.
REFERENCES 2          1. NUREG-0933,      "A Status Report on Unresolved Safety Issues",
U. S. Nuclear Regulatory Commission, April 1989.
l Amendment F A-104                                        December 15, 1989
 
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CESSAR #nincui .
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: 2. 10  CFR  50.61,  " Fracture  Toughness. Requirements for Protection  Against  Pressurized Thermal Shock Events",
Federal Register, July' 23,1985.
i
: 3. American Society of Mechanical Engineers, Boiler & Pressure-                      !
Vessel Code, Section III      (Nuclear),- American Society of Mechanical Engineers.                                                              .
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                                                                                      .s i-l' L
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Amendment F l                                  A-105                December 15, 1989 l
 
.                      CESSAR !!!ninem.                                                                                                                          .
4 B-60    LOOSE PARTS MONITORING SYSTEM
'l ISSUE Generic Safety Issue (GSI) B-60 in NUREG-0933 (Reference 1),
addresses the use of a loose parts monitoring system to detect debris in the reactor coolant. system (RCS) which could damage RCS l
components and/or fuel.
A loose--part - whether it be from an item inadvertently left in the    primary                    system        during          construction,                  refueling,    or
!_                    maintenance, or from component failures - can contribute to further component damage and material wear by frequent impacting with other parts in the system. A loose part can potentially create a partial core flow blockage which                                                    could result in failure of fuel cladding. In addition, a loose part may increase-the potential for control rod jamming and for accumulation of                                                              ^
increased levels.of radioactive crud in the primary system.
The primary purpose of the loose part detection program is the L                      =arly detection of loose matcllic parts in the primary system.                                                                            <
Early detection can provide the time required to avoid or i                      mitigate safety-related damage to,                                    or malfunction of, primary                                          -
system components. Therefore, the NRC established the guidance in i
Regulatory Guide 1.133, Rev. 1 (Reference 2).
SCCEPTANCE CRITERIA l                      The acceptance criterion for the resolution of GSI B-60 is that a plant shall have a loose part monitoring system which is capable of early detection of loose metallic parts in the primary system.
l I
The system should have design features which are identified in Regulatory Guide 1.133 and include at least two acoustic sensors, an appropriate minimum system sensitivity and physical separation of each channel.
'                      In  addition,                    the    system should be                      designed with a data acquisition                    system    with both manual                      and automatic start-up capability, an established " alert level" for loose parts, and capability for sensor channel operability testing.
Finally, the system should be designed for expected environmental and seismic conditions, contain quality components, and provide for enhanced maintainability.
.                                                                                                                          Amendment F A-106                                  December 15, 1989 2 _ _ _ _ _ _ _ _ _            . _ . _ _ _ . _ . __            _ _ _  _ _ _ _ _ . .        _ _ _ .      _ _ _ _              __ ,_  _    ,      . . . _ _
 
C E S S A R h"a ncui.
l BERDLUTION The System 80+ Standard Design utilizes a Loose Parts Monitoring System (LPMS)    to detect the presence of loose parts in the                          i reactor' coolant system (see CESSAR-DC, Section 7. 7.1. 6. 3, also                      I the response to GSI C-12) .      The primary function of the LPMS is                    (
to detect the presence of a loose part within the primary                              l pressure boundary. The secondary function of an LPMS is to i
provide diagnostic information that will assist in determining:                        1 (1) the nature of the loose part(s) (e.g., fixed or free); (2)                        i the location of the loose part; and (3) the characteristics of the loose part (e.g., size, mass,.and velocity). In addition the LPMS is designed to provide the operator with automatic and manual start-up capability.      Furthermore, the system is designed to meet the guidance established in Regulatory Guide 1.133, Rev.
1 (Reference 2).
LPMS sensors are installed at the locations given in CESSAR-DC                    R Table 7.7-4. These locations correspond to natural collection                        1 regions for loose parts'in the primary system and secondary side the steam generator. The two sensors at each natural of collection region and their associated cabling and amplifiers are                    -
physically separated.
Signals    from the sensors are routed via high-temperature,                          a low-noise cable to amplifiers. The amplifier output is L  transmitted to alarm units located within the - control complex.
The alarm unit compares the peak value of the accelerometer output to a predetermined threshold or " alert level"'and provides an alarm to the control room operator via the Data Processing System.-
Finally,    the  LPMS    is  qualified  for  the  expected      normal containment environment and is seismically qualified for an operating basis earthquake. Since the LPMS is important to the
    -safe operation of the plant, limiting conditions for operation
!    (LCO's)  will be provided to the owner-operator in CESSAR-DC L  Chapter 16.      The LCO's for the LPMS are included to ensure L  accurate monitoring of the NSSS for excessive vibration and/or l  debris during plant operation.
Since the loose parts monitoring system meets the intent of l  guidance given in Regulatory Guide 1.133, Rev.1 (Reference 2),
this issue will, therefore, be closed out for the System 80+
Standard Design upon satisfactory completion of NRC review of CESSAR-DC Chapter 16.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
Amendment F A-107                December 15, 1989
 
      -b            I                                                                                                                                            I
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f                      ' 2. Regulatory Guide                  1.133,                  Rev. '1, " Loose-Part                        Detection-p                              Program- for the                  Primary System                  of              Light-Water-cooled i                              Reactors", U.S. Nuclear Regulatory Commission, May 1981.                                                                          1
: i.                                                                                                                                                            'i 5
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Amendment F                  '
A-108                                      December 15, 1989
                            ..      _    - - . - . . . _ . ~ .  . . . _ _ _ . . . ,    . _    _ . . - _ .._.--. ..._ _.. ,---. .              - _ . . -
 
7 CESSAR EH!%an.
C-04:  STATISTICAL METHODS FOR ECCS ANALYSIS ISSUE Generic Safety. Issue (GSI) C-04 in NUREG-0933 (Reference 1),
addresses changes that can be made to the conservative statistical method for ' the Emergency Core Cooling System (ECCS) evaluation model.
Since 1974 the NRC requirements for performing a loss-of-coolant-                          ;
accident - (ICCA)      licensing analyses                  (ECCS analyses)  have been    .
specified in 10 CFR - 50. 4 6 Appendix K (Reference 2). During the                          l years since 1974, extensive research has been conducted on the                              (
    -    various aspects of - a IDCA. Because of this research, 10 CFR 50                            l now- states      that  " ...It        is          now confirmed  that  the methods      J specified in Appendix K,                combined with other analysis methods                i currently      in  use,    are          conservative. and        that  the  actual    j cladding temperature would be much. lower than. that calculated                            ,
using Appendix K methods".                                                                  l l
The NRC has amended 10 CFR 50.46 Appendix K to permit an alternative ECCS analysis method in addition to the conservative approach, to- ECCS analysis. This alternative consists of a realistic ECCS analysis plus an accounting for the uncertainty of the calculation in the adverse direction.                          This method should l        produce a reduced calculated peak clad temperature and would,.
'        therefore be beneficial with respect to plant operation and lifetime. The actual degree of benefit would, however, vary from vender to vendor due to design differences.
ACCEPTA1GE CRITERIA-The acceptance criterion for the resolution of GSI C-4 is that the plant designer must use and meet one of the two ECCS evaluation methods described in 10 CFR 50.46 (as stated above).
RESOLUTION The ECCS design for the System 80+ Standard Design has been improved.      For example, the design uses direct vessel injection and four high pressure safety injection pumps (HPSI) pumps while previous plants typically have cold leg injection and two HPSI pumps. These and other design changes preserve margin for large LOCAs      and significantly                improve      margin  for  small  LOCAs.
Therefore, the performance margin for the System 80+ Standard Design based upon models addressing 10 CFR 50.46 Appendix K l
l          requirements has been significantly increased.
1' l
Amendment F A-109                December 15, 1989
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CESSAR Un%uio.
In addition, the present ECCS model preserves the conservatism upon which fuel design criteria and operational requirements were based for a previous design. Therefore, the conservative ECCS evaluation model identified in 10 CFR 50.46 Appendix K, which has been approved by the NRC, is used for ECCS . analysis for the System 80+ Standard Design.
Since ths conservative ECCS evaluation model' identified in 10 CFR 50.46 Appendix K, remains valid for LOCA analysis, and since this method is approved by the NRC, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
2.- 10 CFR 50. 46 Appendix K, "ECCS Evaluation Models", Code Of Federal Regulations,        Office of the  Federal Register, National Archives and Records Administration.
Amendment F A-110            December 15, 1989
 
CESSAR Mne..
C-05t                                DECAY WRAT UPDATE ISSUE Generic Safety Issue (GSI) C-05 in NUREG-0933                                      (Reference 1),
addresses the need for a designer to select a specific decay heat function for the nuclear power plant LOCA analysis. There are two permissible decay heat functions,.with associated uncertainties, which can be used in ECCS evaluation models. Each of these decay host functions is a part of a particular LOCA evaluation model.
In 1974,  the NRC documented in 10 CFR 50.46, Appendix K, (Reference 2) the required features of LOCA evaluation models.
10 CFR 50.46,-Appendix K specifies that the fission product decay heat generation    function be                                  based on ANS Standard 5.0,    '
(Proposed), (Reference 3) plus a 20% uncertainty factor.
More recently,    a  " Summary                              of    Rule Changes" (Reference  4),
promulgated by the NRC, amended 10 CFR 50.46, Appendix K,                                    such    ;
that ECCS analysis can be performed by either of two approaches.                                      -
The -historic conservative approach may continue to be used - that specifies the ANS ' Standard 5.0 decay heat function, plus 20%
uncertainty as a part of the current I4CA analytical methodology.
This approach is the same as the treatment of decay heat in LOCA analyses historically performed to demonstrate compliance with 10 CFR 50.46.
The alternative approach is to replace the highly conservative method of 10 CFR 50.46, Appendix K, by a realistic analysis plus an accounting for the uncertainty of the calculation in the-adverse direction.        An acceptable function for the realistic fission product decay heat is provided in ANS Standard 5.1, (Reference 5). The acceptability of this function has been identified in NUREG-0933.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI C-05 is that the plant designer shall select either the historic, conservative approach for predictions of decay heat employing the function described in ANS 5.0 (Proposed) or the recent, more realistic approach using the fission product decay heat function identified in ANS 5.1, in order to meet the requirements of 10 CFR 50.46.
Amendment F A-111                December 15, 1989
 
r                                                                                  ;
1 CESSAR aniinemon RESOLUTION The System 80+ Standard Design addresses loss-of-coolant (LOCA).
events - by incorporating a conservative LOCA analysis with a conservative system design.
With respect _ to conservative LOCA analysis, the System 80+                    '
Standard Design r. sets the criteria of 10 CFR 50.46, by using the highly      conservative    fission  product. decay    heat  function        '
identified in ANS 5.0, (Proposed), in lieu of the new and more realistic. decay heat model described in ANS 5.1~.
In addition from a design standpoint, the Safety Injection System has been improved by including two additional high pressure safety injection pumps (for a total of four pumps) .          Because of        i this upgrade, the safety margin for the LOCA analysis for the System 80+ Standard Design has been increased.
Because of conservative system design and analysis for LOCA events the requirements of 10 CFR 50.46 are met and this issue is            -
resolved for the System 80+ Standard Design.
REFERENCES 1._      NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
!    2.      10 CFR 50.46 Appendix K, "ECCS Evaluation Models", Code of Federal Regulations,    Office  of  the  Federal  Register, National Archives and Records Administration.                          ,
: 3.      American Nuclear Society Standard (Proposed) ANS 5.0, " Decay l            . Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors", October, 1971.
: 4.      Summary of Rule Changes, Section 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors, Federal    Register,    Vol. 53,    No. 180,    P.36000,
(.            September 16, 1988.
1 l
l    S.      ANSI /ANS 5.1, " Decay Heat Power In Light Water Reactors",
:              American National Standards Institute, 1979.
1 I
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Amendment F              l A-112                December 15, 1989        ;
l 1
 
CESSAR !!ntricci.
l l                        C-12              PRIMARY SYSTEM VIBRATION ASSESSMENT I
ISSUE Generic Safety Issue (GSI) C-12 in NUREG-0933 (Reference 1) ',
addresses the potential adverse effects of vibration in the Nuclear -- Steam Supply System (NSSS) and the means. of monitoring for. vibration.
Concerns have been expressed about damage to primary systems and components as the result of excessive vibration. A major source of vibration for the NSSS is flow-induced vibration (i.e., water flowing through the Reactor Coolant System (RCS)). -Flow-induced L
l vibration can. lead to damage to the reactor vessel internals and, t
potentially, interference with control rod movement.
The safety concern with respect to vibration is that excessive-                                                                _
wear of the. primary system components could lead to the premature                                                              7  '
failure of those components and to the subsequent release of                                                                    ,
debris into the RCS where it could damage other components (,e.g.,
the fuel, instrumentation, control rods), or to release of radioactivity (which might occur, for example, during a steam generator tube rupture event).                                                                                                      -
KCCEPTANCE CRITERION The acceptance criterion . for the resolution of GSI C-12 is that the system design shall include an effective method for monitoring and controlling system and/or component vibraticn.
This shall' include meeting the intent of guidance provided in SRP Section 3.9.2, Rev. 2 (Reference 2), and Regulatory Guide 1.133 Rev. 1 (Reference 3) . ' (See also GSI B-60 which-has requirements for an vibration monitoring system.)
Specifically, the design requirements of 10 CFR 50, Appendix A (Reference 4), (GDC'S 1, 2, 4, 14 and 15) shall be met.
j Furthermore , specific acceptance criteria concerning the above
!          GDC'S can be found in SRP Section 3.9.2, Rev 2, Section II.
RESOLUTION The NSSS for the System 80+ Standard Design addresses the problem of vibration in two ways. First, by consideration of vibration during the design phase, and, second, by monitoring vibration during plant startup and operation. Experience from the startup of the System 80 plants at Palo Verde is included in the System 80+ Standard Design.                SRP Section 3.9.2, Rev. 2. and Regulatory Guide 1.133, Rev. 1 guidance is incorporated into the system and component design process (see CESSAR-DC, Chapter 3, Sections 3.1 and 3.9.2).
Amendment F A-113                            December 15, 1989
 
CESSAR niWi..m.,
I The GDC'S are specifically addressed in Section 3.1 of CESSAR-DC.
!          Items such as pre-operational vibration and dynamic                                                            effects testing    on              piping,    seismic                    qualification              testing                    of safety-related mechanical equipment and dynamic system analysis methods  for              reactor  vessel            internals              are  considered                          -(See          "
          'CESSAR-DC, Chapter 3, Section 3.9.2).
The - System 80+ Standard Design includes a vibration and leak monitoring          system        to  monitor            the            integrity          of      the                NSSS (CESSAR-DC,            Section    7.7). This system is called the NSSS Integrity Monitoring System (NIMS).                                      NIMS consists of three subsystens: 1) Internals Vibration Monitoring System (IVMS), 2)
Acoustic Leak Monitoring System                                  (ALMS),      and    3)      Loose                  Parts Monitoring System (LPMS).
The primary function of the IVMS is to provide data from which changes in the motion of the reactor vessel internals can be detected.      The LPMS is designed to detect the presence of a loose part within the reactor coolant system pressure boundary and is l
designed to meet-the guidance of Regulatory Guide 1.133, Rev. 1.
Further detail's ~ regarding these subsystems of NIMS may be found in CESSAR-DC, Section 7.7.1.6 and in the resolution of GSI B-60.
The - AIMS is used to detect leaks from the RCS. at specific locations and therefore does not apply to this issue but is addressed in GSI I.D.5 (3).
In summary,              instrumentation necessary to _ monitor system and component vibration has been considered as part of the design and operation of' System 80+. This-instrumentation meets the intent of SRP Section 3.9.2 and Regulatory Guide 1.133, Rev. 1 and meets the requirements of 10 CFR 50-Appendix A (GDC'S 1,                                          2,  4 and 14).
l Therefore, this issue is resolved for the System 80+ Standard j-          Design.
REFERENCES
: 1. NUREG-0933,            "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April, 1989.
: 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-- LWR Edition",
U.S. Nuclear Regulatory Commission.
l          3. Regulatory Guide 1.133 Rev. 1, " Loose-Part Detection Program L
for the Primary System of Light-Water-Cooled Reactors", May 1981.
: 4. 10 CFR 50, Appendix A, " General Design Criteria", Office of the    Federal            Register,          National              Archives and Records Administration.
l Amendment F A-114                                        December 15, 1989
 
CESSAR !!nincum i
HF 1.3.4t MAN-MACHINE INTE M QJa i
ISSUE Human- Factors      Issue HF 1.3.4 in NUREG-0985                                (Reference 1),
addresses the need to appropriately configure several aspects of                                                  !
the man-machine interface design to reduce the potential for                                                      >
human errors during normal and off-normal operations.                                            These            .
i aspects are:
(a)    local control stations, (b)    annunciator systems, (c)    operational aids, (d)    automation and/or artificial intelligence systems, and (e)  computers and computer display technology.
ACCEPTANCE CRITERIA                                                                                        g The acceptance criteria for the resolution of Human Factors Issue HF 1.3.4 are that:
(a)    each local control- station shall be designed to meet the intent of the guidance given in References ,2,                            3,  4,    and 5; (b)  annunciator systems shall be designed to incorporate the criteria in references 2 and 3                              , and meot the intent of                  ;
References 5, 6 and 7;                                                                                a (c)  operational aids shall be designed to meet the intent of the                                                i guidance given in References 8 through 13; (d)  automatic systems are required to initiate and control all protective actions such that the control room operator is not required to take any action before plant conditions are such that manual action is permitted (IEEE Standard 603, Reference 14);
(e)  computers and computer displays in the control room shall be designed to meet the intent of References 5, 15, and 16.
RESOLUTION
    'The System 80+ Standard Design incorporates a NUPLEX 80+ Advanced Contrcl- Complex (see CESSAR-DC, Chapter 18).                                  Details of the NUPLEX 80+ design relevant to the resolution of HF 1.3.4 are as follows:
(a)  All aspects of the local control stations in NUPLEX 80+ are designed to meet the intent of                                  the guidance given in Amendment F A-115                                December 15, 1989                    ]
l l
l 4
 
i 1
l
        .CESSAR1!nho,.
i l
l References 2, 3, 4,- and 5.                                            The man-machine interfaces                                                        I I
                    -at the local control stations are consistent with- the information presentation and control methodologies used in                                                                                                (
the NUPLEX 80+ main control room.                                                                                                                      ]
  '                                                                                                                                                                            \
The design philosophy of the NUPLEX 80+ local. control                                                                                              '!
stations is described in CESSAR-DC, Section 18.7.1.6.2.                                                                                            .I Adequate communications .are provided between the- local                                                                                                ]
stations and the main control- room as discussed- in                                                                                                    .
CESSAR-DC, Section 9.5.2. Because the actuation of local                                                                                                !
controls is on- a single component basis, indication of locally repositioned components is provided in the main                                                                                                I control room'.                A detailed discussion' of abnormal component                                                                              l conditions which are indicated:by various alarms is given in                                                                                            '
CESSAR-DC, Section 18. 7.1. 6.2.10.                                        It should be nored that                                                      !
in the NUPLEX 80+ design, the ability to achieve cold shutdown during conditions of control room evacuation is provided at the remote shutdown panel.                                                                    Local control D                    stations              are      used- only                        for    maintenance and testing L                    activities. Consistent information presentation and control                                                                                            I
                    , techniques reinforce desired operator performance behavior                                                                                              I and reduce the chance of error during normal and off-normal                                                                                              I operation situations,                                                                                                                              j
          ~ (b)      The NUPLEX 80+ annunciator system meets the intent of the guidance and each of the basic functional criteria given in                                                                                              i References-5 and 6. -The annunciator system is described in 2
CESSAR-DC,              Sections 18.7.1.1.4 and 18.7.1.5.                                                                  Of ' major          ~
importance is the reduction of the stimulus overload which can occur during major transients.                                          This reduction has been achieved by decreasing the number of alarm displays by using                                                                                            i group alarm tiles with dynamic message windows and by                                                                                                .
l including processing algorithms to generate the alarms.
Stimulus overload 'is further reduced by basing alarms on validated parameters instead of on                                                          individual ' sensor                                          j channels. Mode and equipment status dependency are included in the alarm logic to eliminate nuisance alarms. The alarms are functionally grouped (see CESSAR-DC, Sections 18.7.3.2.3                                                                                        -'
and 18.7.3.2.4).                                  Also incorporated into the annunciator system are prioritization; availability of first-out alarm information via the CRT's; implementation of the dark-board concept;              and adherence to the accepted criteria                                                                              for labeling, location, auditory signal intensity, flash rates and        readability.                                    The  appropriate          recommendations                                      in Reference 7 have also been incorporated into the NUPLEX 80+
annunciator system, i
Amendment F A-116                                    December 15, 1989
  .. .      -              _ -    . . _ .    , . _ _ _ . . _ . - _ . - - . -                    . _ . _ _ _ _ _ - _ _ . - _ _ . _ _ . . . _ - . . . ~ . . -
 
CESSARim%mo.
I (c) - The 'NUPLEX . 80+~ man-machine interface employs operator aids primarily to, process data prior to presentation to the control room. operators.            The aids are integrated into the presentation hierarchy through application programs of the Data: Processing System (DPS) and the Discrete Indication and Alarm System _(DIAS).          Each of these systems conforms.to the human factors criteria given in CESSAR-DC, Section 18.7.1.1.                      .
Conformance of NUPLEX 80+ to References 12 and 13 is described in CESSAR-DC, Sections 7.5.1.1.5, 7.5.2.5, and                            l 7.1.2.21.                                                                            i The ' following operator aids are provided as part of- the                          a
,          NUPLEX 80+ man-machine interface (with the corresponding                            ;
: i.          CESSAR-DC Sections indicated).                                                      a i
(1)  Signal    reduction        and  validation  -
18.7.1.4  and            l' l
18.7.3.2.1.6,                                                      ,_
(2)  Integrated Process. Status Overview (IPS0) - 18.7.1.2,.            C (3)  Alarm handling - 18.7.1.5 and 18.7.2.3,                                [
(4)  Critical function monitoring - 18.7.1.8.2 and 7.7.1.10,                ;
(5)  Success path monitoring - 18.7.1.8.2,                                    ,
(6), Core limit monitoring - 7.7.1.8.1, and (7)  Computer aided surveillance testing - 7.7.1.8.2.M.                            l (d)- The control. automation of safety systems in NUPLEX 80+                        t    ;
conforms to the requirements of Reference 14, that is, the                  .;
U '.
automatic systems are designed to initiate and control all                    i      l
(
protective actions such' that ' the control room operator is                  7 not required to take any action before plant conditions are                      .;
such that manual action is permitted (see CESSAR-DC, Section-                      1 L            7.1.2.13). 'The -level of control automation for other i            systems is determined - by the functional allocation of the l            task analysis which is described in CESSAR-DC, Section 18.5.
L            NUPLEX 80+ controls for safe shutdown systems are discussed in CESSAR-DC, Sections 7.4.1 and 7.4.2.              The human factors l            related to the Engineered Safety Features Actuation Systems                        ,
l          =and to automatic controls are discussed in CESSAR-DC, Sections      18.7.1.5        and    18.7.1.6.2.6,    wherein    the acceptability of the automatic controls                for safety and non-safety systems is demonstrated.            Automation  in process control- systems and non-safety component controls are discussed in CESSAR-DC, Sections 7.7.1.1 and 7.7.1.2.              The            !
Megawatt Demand Setter is discussed in CESSAR-DC, Sections l            7.7.1.1.3 and - 7 . 7 .1. 2 . 3 .  'NUPLEX 80+ employs no artificial                ,
L            intelligence systems.
{
;      (e)  The    philosophy      of    information    presentation    and  the            i employment of computer technology in plant operations are discussed in CESSAR-DC, Section 18.7.              NUPLEX 80+ utilizes              ,
L Amendment F                1 l                                            A-117                  December 15, 1989            1 1:
L.
 
a CESSAR nutricanon i                            the computer's ability to process raw data and to manipulate and arrange information to support efficient data access by.
the operator.. Process information is made available in a logically structured hierarchical format which is based on                                      i the - results of functional task analysis.                This format is designed - to.. support monitoring, diagnostics and control tasks.:
The Integrated Process- Status              Overview (IPSO)            is    a dynamically      updated      computer    display    which presents                            i information to the operator to enable assessment of the overall plant process performance.            IPSO has been found to
    >                          improve operator performance- during transients based upon validation    experiments      conducted    at  the      Halden  Reactor                      .
Project. NUPLEX 80+ meets the intent of the human f actors criteria identified in References 15 and 16.
Since all the~ acceptance criteria have been met, the man-machine                                  ._
interface issue.is resolved _for the System 80+ Standard Design.
REFERENCES
: 1.        NUREG-0985,. Rev.      02,  "U.S. Nuclear Regulatory Commission Human Factors Program Plan", April 1986.
: 2.        VanCott & Kincade, " Human Engineering Design for Equipment Design", 1972.
: 3.        MIL-STD-1472C,      " Human      Engineering    Design      Criteria      for Military Systems, Equipment & Facilities", December 1974.
: 4.        NUREG/CR-3696, " Potential Human Factors Deficiencies in the L                              Design of Local Control Stations and Operator Interfaces in Nuclear Power Plants", April l1984.                                                                .
1
: 5.        NUREG-0700, " Guidelines for Control Room Design Reviews'',              _
U.S. Nuclear Regulatory Commission, September 1981.
: 6.        NUREG/CR-3217,    "Near-Term Improvements for Nuclear Power Plant Control      Room    Annunciator      Systems",      U.S. Nuclear Regulatory Commission, April 1983.
l                    7.      .NUREG/CR-3987,      " Computerized Alarm Systems",            U.S. Nuclear Regulatory Commission, June 1985.                                                                  l l
: 8.        NUREG-0696,    " Functional Criteria for Emergency Response                                        l Facilities",    U.S. Nuclear Regulatory Commission, February                                      l 1981.
l' l
Amendment F A-118                      December 15, 1989 1,
~            ..    .    .- . -            - .            .  -.            -          -_            --                    ..
 
I s
CESSAR1!nhi
: 9.  -NUREG-0737,    Supplement 1,        " Requirements for Emergency              -
Response . Capability", _ Generic Letter 82-83, U.S. Nuclear Regulatory Commission, December 1982.
: 10. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
: 11. Regulatory    Guide      1.29,    Rev. 03,    " Seismic  Design Classification",      U.S. Nuclear    Regulatory    ' Commission, September 1978.-                                                  -
: 12. Regulatory Guide        1.97,    Rev. 03,    " Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant and                  =
Environs Conditions During and Following an Accident", U.S.
Nuclear Regulatory Commission, May 1983.
: 13. Regulatory Guide 1.47, " Bypassed and Inoperable Status                  A Indication for Nuclear Power Plant Safety Systems", U.S.                  S Nuclear. Regulatory Commission, May 1973.
: 14. IEEE Standard-603,        "IEEE Standard Criteria for Safety                    ,
Systems'for Nuclear Power Generating Stations", 1980.
: 15. EPRI    NP-3701,-      " Computer-Generated        Display    System Guidelines", Volumes 1 & 2, September 1984.
: 16. NUREG/CR-4221,    " Human    Engineering    Guidelines    for    the Evaluation and Assessment of Video- Display Units",                U.S.
Nuclear Regulatory Commission, July 1985.
Amendment F A-119                    December 15, 1989
 
CESSARunhu.
EF 5.11        LOCAL CONTROL STATIONS
[    ISSUE Generic Safety Issue (GSI) HF 5.1 in NUREG-0933. (Reference 1),
i addresses additional NRC guidance for the-design of local control stations.
ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI HF 5.1 are encompassed in GSI HF 1.3.4.
RESOLUTION
:    The resolution for GSI HF 5.1 is identified in GSI HF 1.3.4 and
;. is. addressed and resolved-in this Appendix.
Since GSI HF . 5.1 is subsumed by the above GSI, this issue is resolved for the System 80+ Standard Design.
f-  REFERENCES
: 1.                  "A Status Report On Unresolved Safety Issues",
NUREG-0933, U.S. Nuclear Regulatory Commission, April 1989.
i i'
4
)
l l
.                                                                                                                              l l
l                                                                            Amendment F A-120                            December 15, 1989 l
 
l 1
u-      ,.
CESSAR1!n%uion l
l RF 5.2: REVIEW OF CRl' FERIA FOR HUMAN FACTORS ASPECTS.
OF ADVANCED INSTRUMEN'fATION AND CONTROLS (ANNUNCIATORS)
ISSUE Generic Safety Issue (GSI) HF 5.2 in NUREG-0933 (Reference 1),
addresses additional NRC guidance for the design of advanced instrumentation and controls, in particular with respect to plant                                                                            i annunciators.
ACCEPTANCE CRITERIA The - acceptance criterik for the resolution of GSI' HF 5.2 are encompassed in GSI HF 1.3.4.
BEROLUTION                                                                                                                                    ,
The resolution for GSI HF 5.2 is included in the resolution for GSI HF 1.3.4 contained in this Appendix.
Since GSI HF 5.2 is subsumed by the above GSI, this issue is resolved for the System 80+ Standard Design,                                                                                                i REFERENCES                                                                                                                                ,
: 1.      NUREG 0933,      "A Status Report On Unresolved Safety Issues",
U.S. Nuclear Regu.latory Commission, April 1989.
1 j
l l
1 l
Amendment F                            l A-121                                            December 15, 1989
 
l'                      ,CESSARannemo, l
l I.C.11          SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURES REVISION w
                                                                                                                          ~
ISSUE Generic Safety Issue (GSI) I.C.1 in NUREG-0933 (Reference 1),
addresses the need for improvement in the quality of operational l-information provided to plant operations and staff personnel in order to enhance normal plant operation and the- prevention and mitigation of plant transients or accidents, Following the Three Mile Island Unit 2 (TMI-2) accident, new guidance was - established to improve the quality of operational information for dealing with emergency events. The objective of the guidance identified in NUREG-0737 and supplemented by Generic Letter 82-33 (References 2 and 3), is to improve the quality of procedures to provide greater assurance that operator and staff actions are technically correct, by making them explicit . and easily understood.for normal, transient and accident conditions.
The overall content, wording, and format of procedures that affect plant operation, administration, maintenance, testing and surveillance are to be evaluatad by the NRC in accordance with NUREG-0737 and Generic Letter 82-33.
  ;                    ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI I.C.1, is that the  intent. of .the guidance identified in NUREG-0737,                                as supplemented by Generic Letter 82-33, shall be met.
Specifically, the guidance is divided into four parts. Parts 1, 2
.                      and 3 involve analyses and preparation of guidelines for the L.
l preparation of emergency operating procedures for small break LOCAs, recognition and prevention of impending core uncovery, and operation of a plant in natural circulation.
The fourth part of this guidance addresses NRC review of L                      procedures, guidelines, and the supporting analyses of various i                      transients. Thus, item 4 is not applicable to the plant designer l                      or the owner-operator.
RESOLUTION The owner-operator of a nuclear power generating facility using a System 80+ Standard Design must meet the intent of NUREG-0737, Supplement 1, as supplemented by Generic Letter 82-33, by establishing emergency procedures which address the evaluation and development of procedures for transients and accidents.
i
  ~
Amendment F A-122                          December 15, 1989
 
CESSAR nuincuior b
Specifically, parts 1, 2, &                            3 listed.in the ACCEPTANCE CRITERIA section;above are to be met.
The ultimate responsibility for meeting NUREG-0737, Supplement 1 and        Generic                    Letter      GL  82-33,          remains with the utility owner-operator.                            Combustion    Engineering, however, assists the owner-operator in establishing _these procedures and. training the plant operators and staff by providing Emergency Procedure Guidelines as contained in report CEN-152 (See Reference 4).
Specifically, Section 1.3 of-CEN-152 addresses the guidance and responses to NUREG-0737, including loss of instrumentation, multiple and consequential failures, adequacy ' of core cooling, operator errors during long-term cooling, and optimal recovery guidelines for other-plant accidents.
Combustion Engineering provides analyses and guidance (CEN-152) to      assist            the            owner-operator        in  meeting      the  guidance        o f-  2 NUREG-0737, Supplement 1 and Generic Letter-GL 82-33. Therefore,                                                  y this issue is resolved for the System 80+ Standard Design.
arrmanycas
: 1.        NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.        NUREG-0737, " Clarification of TMI Action Plan Requirements,"                                            !
U.S. Nuclear Regulatory Cor. mission, February 1983.
: 3.        Generic Letter 82-33, " Supplement 1 to NUREG-0737 - Require-ments 'for Emergency                      Response          Capability".      U.S.      Nuclear Regulatory Commission, December 1982.                                                                      ,
l
: 4.        CEN-152,                    " Emergency    Procedure        Guidelines",              Combustion Engineering, Inc.
l Amendment F A-123                          December 1b, 1989 I
  . - _ .    ..~.._.      . . _ _ _ _ _ . _ . _ _            _  _    _ _ _            _ _ _ ,        _, __.
 
                                                                                                                                          )
1 CESSAR iH#,emo i
l I.D.5 f2):      IMPROVED CONTROL ROOM INSTRUMENTATION -
PLANT STATUS AND POST-ACCIDENT MONITORING ISSUE Generic Safety Issue'(GSI) I.D.5 (2) in' NUREG-0933 (Reference 1),
addresses the need to improve.the operators' ability to prevent,                                                                  ;
diagnose-and properly respond to accidents.
i This issue was ' originally identified .in the TMI Action Plan and resulted in the establishment of new NRC (Reference 2) Guidance requirements.                  for addressing the issue is provided in                                                          :
L. Regulatory Guide 1.47 (Reference 3) which provides an acceptable                                                                  >
lL method for implementing the requirements of IEEE 279-1971 and 10 CFR.50, Appendix B (Criterien XIV) with respect to the bypass or inoperable status of safety systems, and Regulatory Guide 1.97
      -(R5?arence 4) which defines an acceptable method for implementing                                                              _
NRC. requirements to provide instrumentation and to monitor plant variables and systems during and following an. accident.
1'.
l.
McEPTANCE CRITERIA s
The acceptance criteria-for the resolution of GSI I.D.5 (2) are contained in:
: 1. Regulatory Guide 1,47          for emergency safeguards features (ESF)      status      monitoring.              Automatic                            bypassed                or inoperable status indication                  at the                          system level                  is recommended .for the            plant      protection                            system,                  safety l
L            systems actuated or controlled by the protection system
).          and      their    auxiliary    and    supporting                              systems.                  These indications should be provided in the control room and should have manuu.t input capability.
l
: 2. Regulatory      Guide      1.97    for-          post-accident                                  monitoring instrumentation.          This      Regulatory                                Guide                identifies              i criteria        for      design      and              qualification                                of      the instrumentation divided into three categories, designated 1,    2  and    3,  which provide a graded approach to requirements based on the              importance to safety of the variable being monitored.              Criteria                          exist for equipment qualification,          redundancy,          power                          sources,                channel i
availability, quality assurance, display and recording, range,      equipment identification, interfaces, servicing, testing      and      calibration,      human                        factors                    and      direct measurement.        The actual variables to be monitored for a pressurized water reactor are tabulated in the guide by type and the instrur.entation design and qualification requirement category (1, 2 or 3) is identified for each variable.
Amendment F A-124                                              December 15, 1989 w              <              ,,      ..y    .,m.  . - - - - . . . - - ,      ,m--.w...,,,        -.        -...,--m
 
I CESSAR Mne.no,.                                                                          j RESOLUTION                                                                              ;
The System 80+ Standard Design provides bypassed or inoperable                          ,
status indication for the Reactor Protective              System (RPS),                '
Engineered Safety Features Actuation System (ESTAS), the systems they control and their auxiliaries or support systems.                  The L          method of conformance        is  summarized    in  CESSAR-DC      Section p          7.1.2.21, anc is consistent with the guidance of Regulatory Guide 1.47. Additional information regarding RPS operating bypasses and trip channel bypasses        is provided in CESSAR-DC Section                      '
7.2.1.1.5. The monitoring of inoperable status of EST components and ESFAS bypasses is described in CESSAR-DC Section 7.3.1.1.1.                          ';
The status of bypasses is indicated in the control room on the Plant    Protection    System    (PPS)    operator's  module and on                    '
approprista alarms        n the Data Processing System (DPS) and Discrete Indication and Alarm System (DIAS).            Inoperability of                -
ESF systems is indicated at the Safety Monitoring Panel on a dedicated ESF Monitoring Section and through the DPS. Manual                    %      C entry capability exists for entry of uninstrumented conditions                  t that affect the availability or performance of a safety system                          '
(e.g., manual valve status, maintenance activities). This is fully described in CESSAR-DC Section 18.7.1.8.2.                                        ;
System    80+  provides    Post-Accident    Monitoring  Instrumentation      -
,            (PANI) in accordance with Regulatory Guide 197.              Indication is i          provided in the control room by three methods.                  A dedicated l          DIAS-channel P provides continuous display              of    Category    1      .
parameters (i.e., variables that require instrumentation with                    _
Category 1 design and qualification) on the Safety Monitoring                            '
Panel. In addition, DIAS-channel N displays Categories 1, 2 and 3 parameters on the control room panels for both normal and                              ,
post-accident operator use.        These indicators validate normal parameter outputt against the PAMI channels to allow operators to use familiar in:-rmation displays during accident situations.
Both DIAS N and P equipment are seismically qualified and isolated from each other.            The third method of obtaining post-accident information is through the DPS cathode-ray tubes (CRTs), which provide all Regulatory Guide 1.97 parameters. More details and a list of PAMI parameters are provided in CESSAR-DC Section 7.5.1.1.5      on  System 80+      post-accident monitoring.
CESSAR-DC    Section 18.7.1.8      discusses    the    integration      of safety-related information, including PAMI parameters, into the control room for optimum human performance.
The System 80+ control Woom meets the guidance identified in Regulatory    Guides 1.47 and 1.97 for improved plant status indication  and post-accident monitering instrumentation as described above.      Thereford, this issue is resolved for the System 80+ Standard Design.
Amendment F A-125                  December 15, 1989 i _ - _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . . _ _ _ _ . _ _ _ _ _ . _ _ , _ . _ _ . _ - _ _ . , __
 
I I
CESSAR Wa"icari                                                            !
REFERENCES
                                                                                  )
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",            i U.S. Nuclear Regulatory commission, April 1989.
: 2. NUREG-0660, "NRC Action Plan Developed as a Result of the              1 TMI-2 Accident",    U.S. Nuclear    Regulatory        Commission, i May 1980.
: 3. Regulatory Guide    1.47,  " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems", U.S.
Nuclear Regulatory Commission, May 1973.                              l 4
: 4. Regulatory Guide 1.97, Rev.3, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assera Plant Environs              i conditions During and Following an Accident", U.S. Nuclear Regulatory Commission, May 1983.                                      l I
i i
                                                                                  )
l l
i
                                                                                  ?
                                                                                  )
i i
l 1
l l
I f
Amendment F          i A-126                  December 15, 1989    l l
i l- --
 
CESSAR Mncam.                                                                                                              !
l          I.D.5 f3)                          IMPROYED CONTROL ROOM INSTRUMENTATION -
i                                QM-LINE REACTOR SURVEILLANCE SYSTEM ISSUE                                                                                                                      i I
Generic Safety Issue (GSI) I.D.5 (3) in NUREG-0933 (Reference 1),
addresses the benefit to plant safety and operhtions of continuous on-line automated surveillance systems.
Continuous        on-line                        surveillance      systems          which    automatically monitor reactors can benefit piknt operations and safety by                                                                '
providing diagnostic information which can predict anomalous behavfor and thus be used to maintain safe conditions.
Various methods of on-line reactor surveillance have been used, including neutron noise monitoring in boiling water reactors (swRs)  to detect                              internals vibration,                and pressure noise surveillance at TMI-2 to monitor primary loop degasification.                                                              ;
on-line surveillance data has been used in the assessment of loose thermal shields.
ACCEPTANCE CRITERIA continuous on-line surveillance of the Nuclear Steam Supply                                                                ,
System (NSSS) involves a number of areas for which acceptance criteria are separately defined:
(1)  Vibration monitoring of reactor internals, (2)  Reactor coolant pressure boundary leakage detection, and (3)  Loose parts monitoring.
The acceptance criteria for the resolution of GSI I.D.5 (3) for internals        vibration                            monitoring  are provided in ANSI /ASME OM-5-1981, (Reference 2) .                                  This standard provides non-mandatory recommendations on the use of ex-core neutron detector signals for monitoring of core barrel axial preload loss. This standard also documents a program containing baseline, surveillance and diagnostic          phases                        and    providen  recommendations              for    dita acquisition frequency and analysis.
The acceptance criteria for leak monitoring are provided by                                                              i Regulatory        Guide 1.45 (Reference 3).                                    This guida documents acceptable methods for leakage separation, leakage detection, detector sensitivity and response time, signal calibration and seismic qualification of pressure boundary leakage detection systems.      It also defines the regulatory position for acceptable design of these systems.
Amendment F A-127                          December 15, 1989
 
CESSARiMeni.                                                                                    ;
i The acceptance criteria for loose parts monitoring (loose part detection in the primary system of LWRs) are provided by Regulatory Guide 1.133 (Reference 4). This Regulatory Guide provides    guidelines  on    system          characteristics          such  as sensitivity, channel separation, data acquisition, and seismic and environmental conditions for operability. It also identifies alert levels, data acquisition modes, safety analysis reports and tichnical specifications pertaining to a locse parts monitoring                                ,
system.                                                                                        ,
RESQLUTICM The System 80+ Standard Design incorporates a NSSS Integrity Monitoring System, which is a system that detects deterioration 1
j    of the NSSS pressure boundary. A description of this system is documented in CESSAR-DC, Section 7.7.1. 6. The system has three subsystemst    the Internals vibration Monitoring System (IVMS),
the Acoustic Leak Monitoring System (ALMS), and the Loose Parts Monitoring System (LPMS).
The IVMS is a monitoring system which gen rates data allowing detection of the motion of reactor internals. It uses linear summed detector signals from each of the ex-core neutron flux ch4nnels. The IVMS also has the capability to perform the analyses    recommended  by ANSI /ASME OM-5-1981.                    The  system l    function, theory of operation and description are provided in l    CESSAR-DC, Section 7.7.1.6.1.        (A related generic safety issue (GSI C-12) also addresses the IVMS.)
i The ALMS detects leaks at specific locations or within specific components in the primary coolent system.                This system uses accelerometers to detect the presence of a primary leak. The system follows, in part, the guidance identified in Regulatory Guide 1.45,    as discussed in CESSAR-DC, Section 7.1. 2. 2 0.                The system functions, theory of operation and description are given in CESSAR-DC, Section 7.7.1.6.2.              Other leak detection methods employed in ths System 80+ standard Design are discussed in                                    !
l CESSAR-DC, Section 5.2.5.
Th LPMS detects the presence of loose parts within the primary l
peassure boundary and provides diagnostic information relating to detected loose parts.      The system is designed consistent with guidance provided in Regulatory Guide 1.133 (Reference 4). The LPMS function, theory of operation, and system description are provided in CESSAR-DC, Section 7.7.1.6.3.                (A related generic safety issue (GSI B-60) also addresses the LPMS.)
i Amendment F A-128                        December 15, 1989
 
#                                                                                                                B C E S S A R ti M icau.
l In summary, the .NSSS Integrity Monitoring System as described above meets the guidance of the applicable Regulatory Guides and the requirements of the applicable codes and standards. Thus, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. ANSI /ASME CM-5, " Inservice Monitoring of core Support Barrel Axial Preload in Pressurized Water Reactors", November 1981.
: 3. Regulatory Guide 1.45,    " Reactor Coolant Pressure Boundary Leakage    Detection    Systems",  U.S. Nuclear            Regulatory Commission, May 1973.
: 4. Regulatory Guide    1.133,  Rev. 1,    " Loose Part Detection                            a      ,
Program  for  the  Primary System    of    Light-Water Cooled Reaccors", U.S. Nuclear Regulatory Commission, May 1981.
L I
Amendment F                                      l A-129                  December 15, 1989                                l l
l
 
CESSAR!Recm..                                                                                                                      l 1
I.D.5 (4)t IMPROVED CONTROL ROOM INSTRUMENTATION -                                                                    i PROCESS MONITORING INSTRUMENTATION ISSUE                                                                                                                            :
Generic Safety Issue (GSI) I.D.5 (4) in NUREG-0933 (Reference 1),
addresses the benefit to plant safety and operations of improved measurement of certain reactor parameters (e.g., reactor vessel water level and relief valve flow), and parameters outside of their_ normal operating range.
A    need      to  improve          process        monitoring        instrumentation                    was                  J identified because of the TMI-2 Accident.                            This need resulted in                                      '
the development of new and improved monitoring systems such as inadequate          core          cooling        instrumentation,            extended                  range post-accident monitoring of certain reactor parameters, reactor level monitoring systems                        and,      monitoring      systems                for    the detection of primary pressure boundary leakage.
ACCEPTANCE CRITERI4 The acceptance criteria for the resolution of GSI I.D.5 (4), for improving process instrumentation are provided by NUREG-0660 and NUREG-0737        (References 2 & 3, respectively). Item II.F.2 of NUREG-0737 provides guidance for the design of instrumentation for detection of inadequate core cooling (ICC).                                    Item II.D.3 of                                .
NUREG-0737 provides guidance on direct indication of relief and                                                                '
safety valve position.
The acceptance criteria                      for the extended range sensors                                are provided by Regulatory Guide 1.97 (Reference 4) in a tabulation of acceptable ranges for post-accident monitoring instrumentation (PAMI).
RESOLUTION
,      The System 80+ Standard Design incorporates                                  improved process
!      monitoring          instrumentation which includes                            ICC            monitoring i
instrumentation.                The  design,            which meets        the          intent          of NUREG-0737        Item          II.F.2,      monitors ICC conditions                          through        a combination          of          resistance        temperature        detectors                    (RTDs),
pressurizer pressure sensors, and reactor vessel level monitors
(      that use integral heated junction thermocouples (HJTCs) and core exit thermocouples.                  Sensor information is processed through I      algorithms to                indicate      loss      of    subcooling,          occurrence of u      saturation and achievement of subcooling conditions following core recovery. The reactor vessel level monitors indicate Amendment F A-130                            December 15, 1989
 
l CESSAR !!nincuio,,
1 I
decreasing                  liquid    inventory            in      the  reactor                              vessel  and the      core              exit    thermocouples                  indicate          steam                        temperatures associated with ICC conditions.
ICC monitoring information is provided to the operator through                                                                        i the Discrete Indication and Alarm System (DIAS) Channel N and the Data Processing System (DPS) displays in the main control room.                                                                      l A complete description of the inadequate core cooling monitoring instrumentation is provided in CESSAR-DC, Sections 7.5.1.1.7 and                                                                    j 7.7.1.1.0.
Extended range sensors have also been incorporated into the System 80+ Standard Design for post-accident monitoring. The ranges      of            instrumentation provided are consistent with the guidance            given            in      Regulatory                Guide            1.97.                        CESSAR-DC, Section 7.5.1.1.5                    describes              the        post-accident                                monitoring instrumentation provided, and Table 7.5-3 lists the pAMI channels and      their            ranges.      Analysis            of      the  post-accident                              monitoring _
instrumentatien is given in CESSAR-DC, Section 7.5.2.4.
* Monitoring of reactor coolant system (RCS) safety valve leakage is provided by in-line RTDs upstream of the safety valve headar.                                                                  ,
Positive indication of safety valve leakage consistent with the guidance of NUREG-0737 Item II.D. 3 is provided by the Acoustic Leak Monitoring System                        (ALMS)        through accelerometers mounted                                        ,
downstream of each valve. Control room alarms are actuated if valves are not fully closed.                                Monitoring of safety valves is                                    7 discussed in CESSAR-DC, Section 5.2.5.1.2.1.
The process monitoring design includes instrumentation for the detection of inadequate core cooling, reactor vessel level monitoring, extended range sensors for post-accident monitoring and monitoring of safety valve leakage.                                  The design is consistent with the guidance given in NUREG-0660, NUREG-0737, and Regulatory Guide 1.97, and ,therefore, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1.      NUREG-0933,              "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.      NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2              Accident",      U.S.          Nuclear        Regulatory                              Commission, May 1980.
: 3.      NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, November 1980.
Amendment F A-131                                                December 15, 1989
 
i:                                                                                                            ;
CESSAR timne            1,..
: 4. Regulatory Guide          1.97,      Rev. 3, "Instrunontation for LWR                              ;
Nuclear Power Plants to Assess Plant and Environs conditions During and Following an Accident", U.S.                          Nuclear Regulatory                    !
Commission,.May 1983.                                                                                    s i
4 Amendment F                      '
A-132                              December 15, 1989
 
1 l
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CESSAR m#icu.
II.B.11        unCTOR COOLANT SYSTEM VENTS issue Generic Safety Issue (GSI) II.B.1 in NUREG-0933 (Reference 1),                                                                                I addresses        the      requirements                                  in      10  CTR  50  and NUREG-0737                                I l
(References 2 and 3,                    respectively) to install reactor coolant system (RCS) and reactor vessel high point vents.                                                                                              l l
After the TMI accident the NRC deterinined that there was a need                                                                              I for vents .in the high points of the reactor coolant system and reactor vessel.                The purpose of these vents is to release                                                                      .
l      non-condensible gases from the RCS which may inhibit core cooling                                                                            j l
during natural circulation. Since the vents ara part of the reactor coolant pressure boundary, the design of the vents must conform to the requirements of 10 CTR 50, Appendix A.                                                                                        t In addition, the NRC determined that the vents should not cause an unacceptable increase in the probability of a less-of-coolant accident (14CA) , should not challenge containment integrity, and should be designed with sufficient redundancy te assure a icw                                                                        -
probability of inadvertent or irreversible itetuation. -                                                                                      ;
KCCEPTANCE CRITERIA                                                                                                                          ,
The acceptance criterion for the resolution of GSI II.B.1 is that plants shall install reactor coolant system and reactor vessel high        point      vent    systems.                                These      systems  shall              meet the requirements of 10 CTR 50.34 (f) (2) (vi) , 10 CFR 50, Appendix A,                                                                            ,
and the intent of guidance identified in NUREG-0737. In addition, the system (s), shall meet the applicable codes and standards for l,
the RCS pressure boundary.
Specifically,            the RCS and reactor vessel vent systems shall incorporate such design features as high point venting of the RCS and reactor vessel, remote control room operation, positive valve indication (located in the control room) and environmentally and seismically qualified equipment.                                                  Also,  the vents should not cause an . unacceptable increase in the probability of a LOCA, should not challenge containment integrity, and should be designed with sufficient redundancy to assure a low probability of inadvertent or irreversible actuation.
RESOLUTION The System 80+ Standard Design includes a Safety Depressurization System (SDS) which performs the Reactor Coolant Gas Vent (RCGV) function to meet the above requirements.
Amendment F A-133                          December 15, 1989
 
CESSAR nWicui, 1
The RCGV function provides a safety-related means of venting remotely from the control room, non-condensible gases from the                                                                    i reactor vessel upper head and the pressurizer steam space during                                                                  l post-accident conditions (see CESSAR-DC, Section 6.7.1.2.1).                                                                      (
Positive indication of vent isolation valve position is displayed                                                                '
in the control room (see CESSAR-DC, Section 7.5, Table 7.5-2).
The RCGV function design assures that the vents will not cause an unacceptable increase in the probability of a loss-of-coolant accident and should not challenge containment integrity.                                    This is accomplished by the installation of a flow restricting orifice in the vent line which limits flow to less than that of one charging pump    (see. CESSAR-DC, Section 6.7.1.2.3).                            In addition,                  the possibility of inadvertent actuation is minimized because the operator must manually actuate the vent valves from the control room. A complete description of the operation of the RCGV l  function of the SDS is identified in CESSAR-DC, Section 6.7.2.1.1.
Finally,      the  RCGV      function              design            is  seismically                  and environmentally    qualified      for            the    expected        conditions                      as described in      CESSAR-DC,    Section                6.7.1.2.1.        Also,          the      RCGV function design complies with the codes and standards which apply to a system that        is part of the                    reactor coolant pressure boundary.
In summary, the RCGV function of the SDS fulfills the applicable requirements of 10 CFR 50, the guidance identified in NUREG-0737, and the applicable industry codes and standards.                          Therefore, this issue is resolved for the System 80+ Standard Design.
REFERENCE 8
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. 10 CFR 50,      Code of Federal Regulations,                          Office of the Federal    Register,        National                Archives        and            Records l        Administration.
: 3. NUREG 0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission.
l Amendment F A-134                                  Decerber 15, 1989
 
CESSARunho.
II.B.3      Post-ACCIDENT BAMPLING SYSTEM                                                                      j issue Generic Safety Issue (GSI) II.B.3 in NUREG-0933 (Reference 1),
addresses the TMI requirements of 10 CFR 50. 34 ( f) and guidance identified in NUREG-0737 (References 2 and 3, respectively).
Basicelly, plants must install a post-accident sampling system to                                                                                                      !
sample reactor coolant und containment atmosphere.
After the Three Mile Island accident the NRC determined there was a need for improved post-accident reactor coolant and containment atmosphere sampling. This determination was made because the TMI l                                  accident demonstrated that existing sampling system designs were                                                                                                        ;
inadequate for post-accident conditions (e.g., insufficient instrumentation and instrument ranges,                                                                            inadequate                    plant          _
shielding, difficulty in obtaining and processing samples) . The
'                                  purpose of the improved post ~ accident sampling requirements is to                                                                                                .
ensure the provision of a remote, rapid, and safe means to obtain potentially                                            highly    radioactive          samples of both the reactor .
coolant and containment atmosphere after an accident. These samples might then be used to assist plant operators in assessing the                      degree                        of  core    damage,        and          determining          the            level
* of contamination in the containment.
NRC guidance was established in NUREG-0737 to assist nuclear power                              plant licensees                        in  developing                improved      post-accident                            e sampling capabilities                                                  for      reactor                coolant      and        containment atmosphere, i
LCQEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.B.3 is that l
Plants shall modify present reactor coolant and containment atmosphere sampling systems or install new systems to satisfy the applicable post-accident requirements of 10 CFR 50. 34 (f) and implement the guidance identified in NUREG-0737.
The reactor coolant and containment atmosphere sampling systems used for post-accident sampling conditions shall permit sampling of                  the                      reactor        coolant        and  containment                  atmosphere                  without personnel exceeding their individual dose limits. The systems shall permit analyses for radioactive                                                                        noble gases,                      iodine, cesium and nonvolatile isotopes, and also boron and chlorides.
RESOLUTION The System 80+ Standard Design includes the Process                                                                  is designed to Sampling System                                    (see          CESSAR-DC,        Section      9.3.2)            which Amendment F A-135                                  December 15, 1989
 
i CESSAR !!nhu.                                                                                                    !
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i I
1 collect representative samples of liquids and gases in various process systems and deliver them to sample stations for chemical and radiological analyses.
l The system permits sampling during reactor operation, cooldown, and post-accident conditions without requiring access to the                                                    I containment.          Remote samples can be taken of fluids in high radiation areas without requiring access to these areas, thus                                                    I permitting personnel to remain within their radiation exposure limits. The sampling system performs no safety function (See CESSAR-DC, Section 9.3.2.1).
The Process Sampling System design meets the performance criteria described below and further discussed in CESSAR-DC,                                              Section 9.3.2.1.1:
      -      The design meets the intent of guidance in Section II.B.3 of NUREG-0737 and in applicable sections of Regulatory Guide 1.97.                                                                                                    ;
l    -      The design integrates both normal and accident sampling functions,        enhancing operator                  system            familiarity          and shortening sample times.
      -      Periodic functional testing cape.bility is provided to assure system availability and operator familiarity during an accident.
t
      -      The design permits liquid and gaseous sample dilution.
L    -      Collection        and      dilution    of          post-accident                  samples  are performed remotely to the maximum extent feasible.
l
      -      Grab samples are used for laborato::y analyses and on-line l            monitors are used for trends.
      -      Remotely operated valves are powered by assured supplies and have reset features to permit operation after containment isolation.      In addition,          valves located in a potential post-accident environment are environmentally qualified to assure operability.
      -      Two different, non-class 1E, power sources (one from each Permanent Non-Safety-Re'.ated Bus X or Y) are available for post-accident sampling. During a loss of offsite power, an alternate power supply is available to meet the 3 hour recommendation of NUREG-0737 for post-accident sampling and analysis.
Amendment F A-136                                        December 15, 1989
 
i CESSAR !!!#iemo,i
    -    The boron sampling system is available in the event of fire.
In addition, chemical and radiochemical analyses are performed to determine boren concentration, fission and corrosion product              '
activity, crud concentration, dissolved gas and corrosion product concentrations, chloride concentrations, coolant ph, conductivity of the reactor coolant, and noncondensible gas concentration in            ,
the pressurizar.
The configuration of the sampling system is such that, under            -
post-accident conditions, samples of containment liquids and the containment atmosphere are transported to a convenient location for remote grab sampling. (See CESSAR-DC Section 3.3.2.2.1).
* In summary, the Process Sampling System fulfills the applicable            [
requirements of 10 CFR 50.34          (f) and meets the guidance          .
identified in NUREG-0737. Therefore, this issue is resolved for the System 80+ Standard Design.                                        -
REFERENCES
: 1. NUREG=0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.                      '
: 2. 10 CFR 50, Office of the Federal Register, National Archives and Records Administration.
: 3. NUREG 0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission.
l l
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l Amendment F A-137                December 15, 1989
 
1 CESSAR Mem.
II.C.4:  RELIABILITY ENGINEERING IEEE Generic Safety Issue (GSI) II.C.4 in NUREG-0933 (Reference 1),                                              '
addresses the need for a designer and owner-operator developed reliability program            which can evaluate plant safety and reliability.
Industry, (including plant designers and owner-operators) and the NRC are concerned about designing and operating nuclear power plants safely and reliably. Before the advent of Probabilistic Risk Assessment (PRA) it was difficult to systematically assess
(,  plant safety and reliability. Therefore, both industry and I
regulators consider PRA, as part of a comprehensive reliability l    program, to be desirable for future plants. The NRC has placed an                                    :
emphasis on PRA for future plants by                              including        it  in the Standardization Rule (10 CFR 52).
Plant          designers employ a PRA for new plants to identify contributors to severe accident risk, and the accident sequences which are significant. The industry goals for new plant designs include a core damage frequency no great'er than 1.0 x E-5 per year. pRA also provides an analytical tool for evaluating the impact of design modifications on core damage probability and the
!    overall risk to the health and safety of the public. The PRA determines expected system and component availabilities. The plant designer's PRA is a useful tool that can be used by the owner-operator as a basis for a reliability program.                                                        ,
According to NUREG-0933, a reliability program generally includes activities such as determining system availabilities, identifying high component failure                  rates,  determining the causes                      for component failures, and identifying possible corrective actions.                                            ,
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.C.4, is that plant designers or owner-operators shall perform a PRA.
Consistent with the Standardization                  Rule (10 CTR 52), the l    assumptions and results of the PRA must be appropriately addressed in an owner-operator reliability program which incorporates such features as determining system availabilities, identifying high component failure rates, determining basic causes          for  component      failures,    and            identifying        possible l
corrective actions.
L l                                                                              Amendment F L                                              A-138                          December 15, 1989 I'
 
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CESSAR nn%.n 1
RESOLUTION                                                                                                                                    1 As shown in CESSAR-DC, Appendix B, a PRA has been performed for                                                                              l the system 80+ Standard Design and meets the mean core damage                                                                                l frequency goal' of less than 1.0 x E-5 events per year as stated in the Licensing Review Basis Document (LRB) (Reference 2).                                                                                  l The System 80+ Standard Design PRA has two primary purposes. The first purpose, is to identify the dominant contributors to severe                                                                            l accident risk. The second purpose is to provide an analytical tool for evaluating the impact of design modifications on core                                                                                ,
damage probability and the overall risk to the health and safety                                                                              i of the public. This information is then used as input to the                                                                                  i owner-operator reliability assurance program.
In          particular,          the determination of core damage                                                frequency i
attributable to internal events (e.g., LOCAs and Loss Of Offsite                                                                              I Power)                used      the    standard        small-event-tree /large-fault-tree                                          /.^
methodology, with full fault tree linking used for the solution
* of core damage                      event sequences.                                External events such as
* tornadoes and earthquakes are also addressed in the PRA.                                                                  The evaluation of the containment performance er. ployed methodologies consistent with NUREG-1150 (Reference 3). The determination of public risk was based on a calculation of the radic, logical dose at            one-half        mile      from      the              plant                    using      bounding      site characteristics supplied by EPRI. The methodology employed was                                                                            .
consistent                  with    the    methodology                            described in NUREG-2300                            7 (Reference 4). The results of this PRA show that the System 80+                                                                        1~
Standard Design plant meets the industry goal of a mean severe core damage frequency of less than 1.0 x E-5 per reactor year and a mean frequency for occurrence of doses greater than 25 REM beyond one-half mile radius from the reactor of less than 1.0 x E-6 events per reactor year.
PRA assumptions are identified in CESSAR-DC, Appendix B and the owner-operators are required to assure that, these assumptions continue to remain valid for their (plant specific) reliability programs. However, the collection and evaluation of operating data for a plant reliability program are the responsibility of the owner-operator.
In summary, a PRA has been performed for the System 80+ Standard Design (see CESSAR-DC, Appendix B) and meets the requirements of the                Standardization Rule,              (10                    CFR              52).      Requiring      the owner-operator to meet the assumptions of the PRA assures that the PRA remains valid during plant operation.                                                            Therefore, this issue is resolved for the System 80+ Standard Design.                                                                                        j l
l Amendment F A-139                                                December 15, 1989                  l l  -.    .- -_-__ .- .- ..                          .  .. _ _ _ . _ . . - - . . . - . . - - - . .                          -    - - - -
 
CESSAR !!!!?ficui.                                                                  :
l l
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i REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",                    i U.S. Nuclear Regulatory Commission, April 1989.                                  '
: 2. Licensing Review Basis Document transmitted to NRC on August 7, 1989 under CE letter number LD-89-088.
I
: 3. NUREG-1150, " Reactor Risk Reference Document",  U.S. Nuclear Regulatory Commission, January 1987.
: 4. NUREG-2300, "PRA Procedures Guide",  U.S. Nuclear Regulatory Commission, January 1983.                                                      ,
l e
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f                                                                                        I l
                                                                                          \
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1
.                                                                                        l l
l                                                                                        \
l Amendment F A-140                December 15, 1989                ,
I
 
CESSAR !!nineam,.
II.D.it  PERFORMANCE TESTING OF PWR SAFETY AND RELIEF VALVES ZR&D Generic Safety Issue (GSI) II.D.1 in NUREG-0933 - (Reference 1),
addresses the guidance identified in NUREG4737 (Reference 2) for qualification testing to be performed on the block, relief, and safety valves of the Reactor Coolant System (RCS).
After the TMI accident, the NRC datormined that there was a need for performance testing of the RCS block, relief, and safety valves.      This determination was made because the TMI accident                                                        '
demonstrated that these RCS valves may not operate as expected.
The NRC established new guidance in NUREG-0737 which addresses the qualification testing of these valves.                                        Qualification testing of these- valves includes testing based upon both normal and accident conditions.
ACCEPTANCE CRITERIA                                          .
The acceptance criterion for the resolution of GSI II.D.1 is that the qualification testing for the RCS block, relief, and safety valves must be performed in accordance with the guidance                                                                .
l identified in NUREG-0737 and with ASME B&PV code, Section III.
t The performance testing shall include both normal and off-normal
!. (accident) conditions.                        Furthermore, the accident conditions shall be established using the applicable design basis events, including Anticipated Transients Without Scram (ATWS).
RESCLUTION The System 80+ Standard Design utilizes the pressurizer safety valves to protect the RCS from overpressurization as required by the      ASME    B&PV        code,          Section    III  (see                          CESSAR-DC,  Section l.
5.4.13.4.1).
The inlet and outlet portions of the valves are hydrostatically tested with water at the appropriate pressures required by the applicable section of the ASME code.                                    Cet pressure and seat leakage tests are performed with staan using a pro-rated spring.
Final set pressure tests are conducted using the final design l  springs with either high pressure steam or low pressure steam and an assist device.
Tests performed prior to shipment test seat leakage use the final design springs and either hot air or hot nitrogen.                                                            Valve adjustment is made to a valve ring setting combination to provide Amendment F              ]
A-141                                          December 15, 1989 l
l
 
CESSAR !!i#icau.
  , stable valve operation using the EPRI Safety Valve Test Program i  results documented in CEN-227 (Reference 3). This valve test program was based upon the guidance established in NUREG-0737.
The System 80+ Standard Design does not use power operated relief valves (PORV's) and the requirements and guidance regarding these                                                                                                    ,
valves do not therefore                                apply.      (As described in CESSAR-DC,                                                                        ;
Section        6.7,          the    System                80+          Standard            Design                      includes                                  ;
safety-related isolation valves, manually actuated from the
!    control room, to provide depressurization capability.]
NUREG-0737              specifies      that            applicable                    design      basis                          events including ATWS are to be considered in developing performance testing conditions.                The System 80+ Standard Design employs an independent and diverse control-grade reactor trip and turbine trip specifically designed to address the prevention of ATWS                                                                                                          ,
events (see CESSAR-DC, Section 7.7.1.1.11).                                                      This Alternate Protection System (APS) augments the Reactor Protective System for ATWS (see USI A-09 for the resolution to ATWS).                                                                          The APS includes      an Alternate Reactor Trip                                            Signal    (ARTS)                  which                        is            '
separate and diverse from the Plant Protection System.                                                                    The ARTS                                    .
equipment provides a simple,                                        yet              diverse mechanism to                                                            ,
significantly decrease the possibility of an ATWS . - Therefore, there are no special relief valves for the mitigation of ATWS and the corresponding test requirements for GSI II.D.1 do not apply to the System 80+ Standard Design.
Since the testing for pressurizar safety valves conforms to the guidance given in NUREG-0737 and since testing requirements on i
PORV, associated block valves, and ATWS events do not apply, this l~  issue is resolved for the System 80+ Standard Design.
1 REFERENCES
: 1.        NUREG-0933,          "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2.        NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, October 1980.
: 3.        CEN-227, " Summary Report on the Operability of Pressuriter Safety Relief Valves in CE Designed Plants",                                                              Combustion Engineering, Inc, December 1982.
: 4.        Regulatory          Gdide          1.97,          Rev.                3,    " Instrumentation For Light-Water-Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident",                                                                            U.S.
Nuclear Regulatory Commission, May 1983.
Amendment F A-142                                  December 15, 1989
 
I i
CESSAR !!nLm.                                                                                                                                              ;
II.D.31                          DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION ISSUE                                                                                                                                                      ,
Generic Safety Issue (GSI) II.D.3 in NUREG-0933 (Reference 1),
addresses the guidance identified in NUREG-0737 (Reference 2) for the provision of a positive indicati.on in the control room of relief and/or safety valve position or a reliable indication of                                                                                            l flow in the associated discharge piping.
After the Three Mile Island Unit 2                                                (TMI)                      accident,      the NRC determined that there was a need for direct indication of relief                                                                                            :
and                safety              valve  position    in        the                    control                room.        This                      l determination was made because the TMI accident demonstrated l
l that, during an accident-, these valves may not operate as I expected and that the safety and relief valve instrumentation may fail                  to            provide  the  operator    with                    sufficient                    information                e concerning the status of these valves.                                                                  Therefore, the NRC established new guidance in NUREG-0737 which addresses the installation of improved safety and relief valve indication in the control room to enhance the operator's ability to diagnose a safety and relief valve failure and/or incorrect position.                                                                                                  ,
l ACQEPTANCE CRITERIA i
l The acceptance criterion for the resolution of GSI II.D 3 is that                                                                                    .
the plant design shall include safety and relief valve indication                                                                                          :
l in accordance with the guidance given                                                            in NUREG-0737.                  This                      '
indication shall have the following design features
  -                Unambiguous safety and relief valve indication shall be provided to the control room operator;                                                                                                    ,
  -                Valve position should be indicated within the control room and should be alarmed Valve                position    indication    can be                                    either        safety    or control grade.                If it  is control              grade,                      it  must    be powered from a reliable (e.g., battery-backed) instrument bus (see Regulatory Guide 1.97 (Reference 3));
  -                valve position indication should be seismically qualified consistent with the component or system to which it is attachedi
  -                The valve position indication shall be qualified for the l                  appropriate                operating    environment                            which            includes        the expected normal containment environment and an Operating Basis Earthquake; and Amendment F A-143                                                    December 15, 1989
_ . _ . _ _ _ _ ._ _ _ ~ _.. _ ._ _                              _ _ . _ . _ _ . _ _ _ _ _ _ . , _ _ . , . _ _                        _ _ . _ - - - -
 
i CESSAR Memos                                                                                      ,
The  valve  position  indication                    shall  be  human-factors          l engineered.                                                                                ,
gasoLUTIon                                                                                        ;
The System 80+ Standard Design incorporates four primary safety                                    ,
valves (see CESSAR-DC, Section 5.4.1.3).                            Valve discharge is            ,
headered and routed to the In-containment Refueling Water Storage Tank. These valves are monitored by three methods which are described in CESSAR-DC, Sections 5.2.5.1.2.1:
  -      First,  positive indication of safety valve position is supplied in the control room by the Acoustic Leakage                                        :
Monitoring System (ALMS).
  -      Second, each safety valve is monitored for seat leakage by                                  :
an in-line Resistance Temperature Detector (RTD) which is located upstream of the header for the safety valves.
  -      Third, safety valve leakage is indirectly monitored from the safety grade pressurizer pressure and level. instrumentation system also located in the control room.
The ADES is part of the NSSS Integrity Monitoring System and is described in CESSAR-DC, Sections 7.7.1.6.1 and 7.7.1.6.2.                              The function of the ALMS is to detect a leak at specific locations or within specific components in the primary system including the primary safety valves.        The ALMS 'prevides                    the control    room operator with a direct and unambiguous method of determining the position (open or closed) of the pressurizer safety valves as required by NUREG-0737.                                                                            i The ALMS      is composed of sensors (accelerometers) which are installed on the pressurizer safety valve discharge lines (one per safety valve) . Signals from the sensor area are routed to the    in-containment amplifiers.                        The    amplifier    output      is subsequently directed to the control room.                          Within the alarm instrumentation, the signal is compared to a threshold value obtained during startup testing.
Alarms are provided as part of the " human engineered" control complex (see CESSAR-DC, Chapter 18) and are included in the plant computer annunciator systems.                      Af ter passing through the alarm unit,    the amplified    accelerometer                    signals are multiplexed, filtered, digitized, and transmitted to a computer for further analyses.      The computer maintains                        data  storage,    performs comparisons, develops trends, and performs analyses to enhance the signal characteristics.
Amendment F A-144                                December 15, 1989
 
I a
1 CESSAR !!nineuios i
The AIMS is qualified for the expected normal containment                        .
environment and is seismically qualified for an operating Basis                  )
Earthquake. Finally, as identified within Regulatory Guide 1.97,                ;
the AIMS is supplied with power from non-vital buses X or Y.                        1 These buses are very reliable since they use batteries as a backup power source.
In summary, by providing a direct method for monitoring safety valve position, the ALMS implements the guidance identified in                  :
NURIG-0737. Therefore, this issue is resolved for the System 80+
Standard Design.                                                                ;
REFEMMCES                                                                        ,
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",                ,
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG ~J737, " Clarification of TMI Action Plan Requirements'',
U.S. Nuclear Regulatory Commission, June, 1985.                        .
: 3. Regulatory Guide 1.97, Rev.      3, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assass Plant Environs          '
Conditions During and Following an Accident", May 1983.                    ,
t I
1 L
Amendment F A-145                  December 15, 1989
 
l l:
C E S S A R in ancan.
l l-                                                                                                                                                        )
l l
l II.E.1.11                      AUEILIARY FEEDWATER EVALUATION                                                                l ISSUE Generic Safety Issue (GSI) II.E.1.1 in NUREG-0933 (Reference 1),
addresses the TMI-related requirement                                          (10 CFR 50. 34 (f) (1) ,                                            l Reference 2) that                                  all  operating  plants                  and plants under construction re-evaluate thair emergency feedwater (ETW) system                                                                                    ;
designs with respect to reliability and availability.                                                  -
After the Three Mile Island Unit 2 accident the NRC reviewed the auxiliary feedwater system for availability and reliability of components and decay heat removal etpability. In particular, the EFW system was scrutinized with regard to the potential for failure under a variety of loss of main feedwater conditions. The safety concern was that a total loss of feedwater, i.e., loss of both main and emergency feedwater, could result in loss of core                                                                                  ;
cooling. The NRC requested operating plants and plants under construction to review both the reliability and the capability of the EFW system to perform its intended safety function i.e.,                                                    core decay heat removal. The evaluation by the plants was divided into three parts as discussed below.                                  .
Part one consisted of a limited PRA to determine the potential for EFW system failure under various loss-of-main-feedwater transient conditions, with particular emphasis being placed on determining potential failures from human errors, common causes, single-point vulnerabilities, and test and maintenance outages.
This evaluation applies to operating plants and plants under construction and not to advanced or futura plants. Part two was composed of a deterministic review- of the EFW system using the acceptance criteria of SRP Section 10.4.9 and the associated Branch Technical Position (BTP) ASB 10-1. Part three required a re-evaluation of the decay heat removal capability of the EFW system with respect to EFW system flowrate. Parts two and three apply to advanced or future plants.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.E.1.1 is that plants using emergency feedwater systems shall maat the intent of the guidance identified in SRP Section 10.4.9 and BTP                                                                                    )
ASB 10-1 and thus meet the requirements of 10 CFR 50 Appendix A                                                                                    .
(General Design Criteria (GDC)) and 10 CFR 50.34 (f) (1) .                                                                                        )
Specifically, the EFW system shall meet the requirements of GDCs                                                                                  l 2,          4, 5, 19, 34,              44, 45 and 46 with respect to a variety of design criteria.
l Amendment F                                      '
A-146                                December 15, 1989 l
      . . . . . ~ . -_          .._ _ . _ . _ _ . - . _ ,              . _  . , _ . . . .          __ __  ._      _, . . _ . _ . . _ . . _ _
 
CESSAR !!nhe.
These design criteria include: the capability of the system to withstand the effects of earthquakes and missiles; shared systems and components; prompt shutdown of the reactor from the control room; system decay heat heat removal capacity considering a main feedwater    line    break;    redundancy;  reliability;  in-serv;ce inspection; and functional testing.
RESOLUTION The  System    80+    Standard ' Design incorporates an Emergency Feedwater (EFW)      System to provide a reliable and independent safety-related      means    of  supplying  secondary-side,  quality feedwater to the steam generator (s)          for removal of heat and prevention of reactor core uncovery during emergency phases of plant operation.        The EFW system is a dedicated safety-related system which is not used during normal plant operation (see CESSAR-DC, Section 10.4.9).
w The EFW system censists of two separate mechanical trains each            0 aligned to supply its respective steam generator.            Each train consists of a dedicated safety grade storage tank, two EFW pumps          3 (one electric driven and one steam driven), a cavitating venturi to limit the maximum flow to a faulted steam generator, and the associated valves and instrumentation (see CESSAR-DC, Section 10.4.9.2.1).                                                                -
Consistent with its importance to plant safety, the system has design features which meet the requirements of the GDCs                    ;
identified in the Acceptance Criteria and SRP Section 10.4.9 (including BTP ASB 10-1). For exampl6,          the ETW system design includest EFW components that are located in Seismic Category I structures which protect them from the effects of              external missiles; essential components that are designed to account for the environmental effects of flooding, missiles and earthquakes (components and piping necessary to perform the EFW system safety function are designed to Seismic Category I requirements as described in CESSAR-DC, Section 3.7); and manual or automatic (EFAS or APS) initiation from the control room.              CESSAR-DC, Section 10.4.9.1.2        provides a comprehensive discussion of the EFW system design criteria which fully address the GDCs identified in SRP Section 10.4.9. Also, functional and inservice testing are identified in CESSAR-DC, Section 10.4.9.4.
Emergency Feedwater System reliability is addressed by GSI 124,              i The response to GSI 124 demonstrates that the System 80+ Standard Design EFW _ system fulfills the requirements of component and system reliability.      A complete description of the PRA for the System 80+ Standard Design, including the EFW system reliability analysis is given in Appendix 10A of CESSAR-DC.
Amendment F A-147              December 15, 1989
 
l CESSAR !!n*Fieanon l
Finally, the ETW system is designed to provide decay heat removal capability for 8 hours at hot standby and then support an orderly cooldown to shutdown cooling system entry, even if the initiating event is a main fesdwater line break (see CESSAR-DC, Section 10.4.9.3).
Since the guidance identified in NUREG-0737 and SRP Section 10.4.9    (including  BTP          ASB                    10-1) is    considered        and        the requirements of 10 CFR 50.34 (f) (1) and 10 CFR 50, Appendix A are fulfilled, this issue is resolved for the Syster: 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
j U.S. Nuclear Regulatory Commission, April 1989.
: 2. 10 CFR 50. 34 ( f) (1) , " Additional TMI-related Requirements",
Office of the Federal Register, National                                        Archives        and Records Administration, November 30, 1981.
: 3. 10 CFR 50 Appendix A, " General Design Criteria", Office of the    Federal Register, National Archives and Records Administration.
: 4. NUREG-0800, Standard Review Plan, Section 10. 4.9,                                    Rev.        2,
          " Auxiliary Feedwater Reliability",                            U.S.      Nuclear Regulatory Commission.
t Amendment F A-148                            December 15, 1989
 
I l
CESSAR !!nificm l
1 l
1 II.E.1.2t    AUZILIARY FEEDWATER AUTOMATIC INITIATION                                                                        i AND FLOW INDICATION                                                                                    i 185UE                                                                                                                                            l Generic Safety Issue (GSI) II.E.1.2 in NUREG-0933 (Reference 1),
addresses                  the  TMI    requirement                            for      plants    to      install                a control-grade system for automatic initiation of the auxiliary                                                                                    '
feedwater                (AFW)  system. This requirement can be achieved by meeting                the criteria identified in IEEE Standard 279-1971 (Reference 2), (e.g., ti'maly system initiation, single failure criterion, equipment qualification).
After the Three Mile Island Unit 2 accident, the NRC reviewed auxiliary                feedwater    system                designs                  with  respect        to timely initiation, as described in 10 CFR 50, Appendix                                                      A,          (GDC 20),
(Reference 3). Upon completion of the review, the NRC determined                                                                            ;
that new guidance identified in NUREG-0737, (Reference 4) was                                                                                ";
necessary in order to assure a timely start of the AFW system after a design basis event (e.g.,, loss of main feedwater). Among this new guidance was automatic system initiation, environmental and                seismic    equipment              qualification,                      and    single                failure criterion.
An NRC review of IEEE Standard 279, established that the criteria                                                                              '
for Class 1E or safety-related electrical equipment described therein, are acceptable for the resolution of this safety issue.
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.E.1.2, are that plants employing auxiliary feedwater systems shall meet the requirements of 10 CFR 50, Appendix A, (GDC 20) by implementing the guidance identified in NUREG-0737 and the design requirements of IEEE Standard 279-1971 (References 3 & 4, respectively).
Specifically, the auxiliary feedwater system shall incorporate such design features as automatic system initiation, protection from single failure,                      and environmental and seismic equipment qualification.
RESOLUTION The System 80+ Standard Design utilizes a dedicated emergency feedwater (ETW) system to provide an independent safety-related means of supplying secondary-side quality feedwater to the steam generator (s) for removal of heat during emergency phases of plant operation. The EFW system has no operating functions for normal plant operation. (See CESSAR-DC Section 10.4.9).
Amendment F A-149                                  December 15, 1989
 
C E S S A R n!Mne m ..
In addition, the emergency feedwater system instrumentation and controls are part of the engine <tred safety feature (ESF) systems and are subject to the design bases in CESSAR-DC Sections 7.3 and 10.4.9. These design bases address the applicable GDC identified in 10 CFR 50, Appendix A, including GDC 20.                                                      ,
The  EFW    system  is  actuated  automatically  by      an            emergency feedwater actuction signal (EFAS) from the ESF actuation system or by the auxiliary protection system (described in CESSAR-DC Section 7.7) . In addition to this automatic feature, the ETWS can be manually initiated as described in CESSAR-DC Section 10.4.9.
The ESF actuation system is composed of redundant trains A, B, C, and D.      The instrumentation and controls of each train are physically and electrically separate and independent. The ESF actuation system can sustain the loss of an entire train and still provide its required protective action.                  Specific ESF design criteria are addressed for environmental and seismic equipment qualification, single failure criterion, and minimum equipment and system response times.
In  summary, the emergency feedwater system,            including its integral instrumentation and controls, fulfills the applicable requirements of 10 CFR 50, Appendix A by meeting the guidance identified in NUREG-0737 and the design criteria in IEEE 279-1971.
Therefore, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0933,  "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations",          The      Institute                of Electronic and Electrical Engineers.
: 3. 10 CFR 50 Appendix A, " General Design Criteria", Office of the    Federal Register, National Archives and Records Administration.
: 4. NUREG 0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission.
Amendment F A-150                December 15, 1989
 
CESSAR M eau..
II.F.2t    INSTRUMENTATION FOR DETECTION OF IMADEOUATE CORE COOLING ISSUE
                                                                                                                                          ^
Generic Safety Issue (GSI) II.F.2 in NUREG-0933 (Reference 1),
addresses the need              for plants to                    install            improved                    accident monitoring instrumentation for the detection of inadequate core cooling.
The    TMI      accident,    identified          a          need    for          improved                    accident monitoring instrumentation because at the start of an accident, it may be difficult for the operator to immediately evaluate what                                                                      ,
accident        has    occurred          and,    therefore,                    to              determine          the appropriate response.
Plant instrumentation is required to provide indication to the                                                                      ;
control      room operators of certain plant variables during
* accidents. This accident monitoring instrumentation is necessary to provide information required to permit the operator to take pre-planned manual actions to accomplish safe shutdown of the reactor;          determine      whether        the            reactor              trip,                engineered safety-feature systems,                  and  manually            initiated                  safety-related systems are performing their intended functions (i.e., reactivity control, core cooling, maintaining containment integrity): and provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release (i.e., fuel cladding, reactor                                                                        ,
coolant pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred.
The NRC established guidance for improved accident monitoring instrumentation in NUREG-0737 and in Regulatory Guide 1.97, Rev.
3 (References 2 and 3, respectively). The purpose of the guidance i  is to assist owner-operators and designers in developing improved accident monitoring instrumentation.                              As with previous accident monitoring instrumentation, these improved and/or new systems must meet the applicable GDC's identified in 10 CFR 50, Appendix A (Reference 4).
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI                                                        II.F.2, is that    plants        shall          modify        present                accident                      monitoring instrumentation            and/or          provide              new        accident                      monitoring instrumentation          that meets the                      intent        of          NUREG-0737.                  In addition, this new and/or improved instrumentation shall meet the requirements of 10 CFR 50, Appendix A, (GDC's 13, 19, 64) and implement the guidance identified in Regulatory Guide 1.97, Rev.
3 (as related to inadequate core cooling).
Amendment F A-151                                            December 15, 1989
_. _      _.          __        _ . . .      _  . . . . _        _ _ _ _ . . _ . _ _ _ _ - _ _ _ _ . _ _ ~                        _
 
i i
CESSAR !!nincam.                                                            .
i Specifically, the accident monitoring instrumentation shall be              ,
i designed such that the operator will be provided with sufficient            :
information during accident situations to take pre-planned unual actions, and to determine whether safety systems are operating properly. In addition, the instrumentation will also provide sufficient data so that the operator can evaluate the potential            ,
for core uncovery, and gross breach of protective barriers, including  the  resultant  release    of  radioactivity  to  the      ,
environment.
RESOLUTION                                                                  ,
The System 80+ Standard Design utilizes the Nuplex 80+ Control Room that is designed in accordance with the applicable General Design Criteria (GDC) identified in 10 CFR 50, Appendix A (including GDC's 13, 19, 64) (See CESS AR-DC, Section 3.0) .      The      :
Nuplex 80+ Control Room employs an integrated information display hierarchy to present both safety-related and non-safety-related plant data for monitoring and control by the operator (See                  :
CESSAR-DC, Section 7. 5.1) . All information is integrated, (in            ,
accordance with Regulatory Guide 1.97) such that the same instrumentation use( for accident monitoring is also used for normal plant operatian. If an accident scenario develops, this integration allows the operators to diagnose and monitor the event using instruments with which they are the most familiar.
The Nuplex 80+ information systems also include automatic signal
,    validation, through cross-channel data comparison, prior to data presentation or alarm generation. This comparison ensures that the process information displayed to the operator is correct.
Multiple diverse systems are utilized to process and display the data to ensure that infomation processing errors are detected i
and alarmed. This integrated information display hierarchy is            ,
l    composed of the following major elements: Integrated Process l    Status overview (IPS0) Panel, Discrete Indication and Alarm System (DIAS), Data Processing System (DPS), Component Control System (CCS) and operator displays. A further description of l    these systems can be found in CESSAR-DC, Section 7.5.
The Inadequate Core cooling (ICC) monitoring instrumentation is p rt of the Nuplex 80+ Control Room and is designed to meet the I    intent of the guidance identified in NUREG-0737.            The ICC instrumentation and displays provide sufficient information to permit the operator to evaluate the potential for core uncovery, and gross breach of protective barriers, including the resultant release  of    radioactivity  to    the  environment. The  ICC instrumentation is described in detail in CESSAR-DC, Section 7.5.1.1.7  and consists of the        fellowing  sensor package:
resistance temperature detectors (RTD's), pressurizer pressure Amendment F A-152                  December 15, 1989
 
_e    .,::
      . Vl
['
3    ,
  %. :q, CESSAR E!ninemon sensors, ~ and a reactor vessel level monitoring system (RVLMS) employing . heated junction thermocouples                          (HJTC)        and core exit                              ;
thermocouples.
The signals from the - RTD's, unheated thermocouples in the HJTC system, and_ pressure sensors are combined to indicate the loss of                                                            -
sub-cooling,      occurrence of saturation and                            achievement        of      a sub-cooled condition following core.r*covery.
The reactor vessel level monitors provide information to the-
                    - operator on the liquid level inventory in the reactor pressure vessel regions above the fuel alignment plate.                                    The core exit thermocouples        monitor the            increasing                  steam        temperatures associated '    with-  ICC  and  the          decreasing              steam      temperatures
                    ' associated' with recovery from ICC.                    Details of the ICC sensor
                    -design and signal processing can be found in CESSAR-DC, Sections 7P. 5.1.'1. 7.1 and 7. 5.1.1. 7. 2,  respectively.
The ICC parameters are 'incorporatcd into the data processing                                                            }
system (DPS) Critical Function "
                                                                            .toring (CFM) displays and
                    - alarm logic which are described                      -etail in CESSAR-DC Section,                                          -
7.5.1.1.7.3.3~.                                                                                                          ;
The System 80+ Standard Nuple>                  c      Control Room displays both                                        ,
safety and non-safety related 5.                      information and includes data
!:                    used for the detection of f            -
                                                                        . ate core cooling.                  The Nuplex 2
L                    -80+ control Room is designeu                      accordance with the applicable codes, standards and regulations, (10 CFR 50, Appendix A) and                                                            ,
5 meets    the    intent    of  Regulatory            Guide          1. 9 7 ,-  Rev. 3,    and NUREG-0737, as previously described.                      Therefore, this issue is
                    - resolved for the System 80+ Standard Design.
l l-l                    REFERENCES 1.'  NUREG-0933,    "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. NUREG 0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission.
: 3. Regu1Etory      Guide    1.97,          Rev.      3,          " Instrumentation For Light-Water-cooled Nuclear Power Plants To Assess Plant And Environs Conditions During And Following An Accident", U.S.
Nuclear Regulatory Commission, May 1983.
: 4. 10 CFR 50 Appendix A, " General Design Criteria", Office of the  Federal      Register,        National          Archives and Records
                            -Administration.
Amendment F A-153                                      Decerber 15, 1989
 
l CESSAR !!nincam.                                                                                l l
l, I
II.F.3t  INSTRUMENTS FOR MONITORING ACCIDENT CONDITIONS l    ISSUE                                                                                          ]
L    Generic    Safety  Issue-    II.F.3    in    NUREG-0933    (Reference        1),
addresses the adequacy and availability of instrumentation which monitors plant variables and systems during and following an li  accident. Prior to the Three Mile Island (TMI) Accident, nuclear                            ,
power generating stations were equipped with accident monitoring instrumentation using the guidance identified in Regulatory Guide 1.97, Rev. 1 (Reference 2) and ANSI /ANS 4.5 (Reference 3).
l After    the  TMI  accident,      several    concerns    were identified                    ]
regarding the availability and adequacy of instrumentation to                                  1 monitor plant variables and systems during and following an accident (see NUREG 0737 (Reference 4)). Regulatory Guide 1.97 was revised (under Revision 3) (Reference 5) to incorporate new
    . instrumentation as a-result of the TMI experience.
l l-  Regulatory Guide 1.97, Rev.3 describes                a  method which is j:  acceptable to the NRC for complying with the requirements to
!:  provide instrumentation to monitor plant variables and systems                                  .
l during and following an accident.            ANSI /ANS 4. 5 , delineates the criteria for determining the variables to be monitored by the l
control room operator during the course of an accident and during                                .
the long-term stable shutdown phase following an accident.                                      l
                                                                                                . 1 ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.F.3 is that there shall . be instrumentation of sufficient quantity, range, availability and reliability to permit adequate monitoring- of plant variables and systems during and following an accident.
Specifically, the instrumentation shall conform to the guidance                                l given in Regulatory Guide 1.97, Rev. J                and ANSI /ANS 4.5 and                      l should provide sufficient information to the operator for (1)                                  j taking pre-planned manual actions to accomplish safe plant                                      i shutdown;    (2)  determining        whether      the    reactor          trip,              l engineered-safety-feature        systems,    and    manually      initiated safety-related systems are performing their intended safety functions (i.e. , reactivity control, core cooling, maintaining RCS integrity and containment integrity); (3) determining the potential for causing a gross breach of the barriers to                                        l reactor coolant radioactivity release (i.e.,          fuel cladding, pressure boundarv, and containment) and determining if a gross breach has occurred.
Anendment F A-154                    December 15, 1989
 
CESSARI!ninc.m                                                                                        l 1
l L
RESOLUTION                                                                                            l The System      80+ Standard Design incorporates              the Nuplex              80+
Advanced Control Complex (ACC) which includes the Post-Accident Monitoring Instrumentation (PAMI). The PAMI is designed in accordance with the intent of the guidance given in Regulatory                                        i Guide 1.97, Rev. 3.                This instrumentation is itemized in CESSAR-DC, Section 7.5.1.1.5 and Table 7.5-3 which includes the                                      l parameters monitored, the number of sensed channels, sensor                                          i
        ' ranges, indicated range, location, and associated Regulatory                                          .
Guide 1.97 category. Examples of plant parameters monitored are l
RCS pressure,    primary safety valve position,              primary coolant temperature, containment pressure, and site radiation.
h The Nuplex 80+ ACC includes the Main Control Room (MCR) . The MCR design integrates the Safety Parameter Display System (SPDS) function and the PAMI using three methods. The first method of integration uses the Discrete Indication and Alarm System (DIAS)                              a.
i          Channel-P processors and displays which are dedicated to                                          t l          continuously monitor and display Category 1 parameters such as-RCS pressure, containment pressure, and reactor vessel coolant level within the MCR on the Safety Monitoring panel. The second                                  .
integration method includes DIAS Channel-N displays which are                                  i integrated into the MCR for display of Category 1 and 2 PAMI                                        v parameters during both normal operations and accident conditions.                                  . :
These displays include such parameters as core exit temperature,                                    :
emergency feedwater storage tank level, safety injection tank                                  .  -
level, and plant radiation level.              (DIAS Channel-N is isolated from the DIAS Channel-P displays) .            The third integration method employs the Data Processing System                (DPS) which utilizes CRT displays to provide    indication          for  all    Category 1,2 and 3 parameters    including  main        feedwater    flow,  emergency diesel
        -generator status, and RCS radiation level (these CRT displays are isolated from DIAS Channels P and N). This system also provides integrated displays for Critical Safety Functions, Inadequate core Cooling, and other safety related plant parameters.
These instrumentation and information systems, when evaluated together, provide sufficient information to permit the operator to: (1) take pre-planned manual actions to accomplish safe plant shutdown; (2) determine whether the reactor trip, engineered-safety-feature systems, and manually initiated safety systems important to safety are performing their                    intended safety functions; and (3) determine the potential for a gross breach of the barriers to radioactivity release and to determine if a gross breach has occurred. Since these instrumentation and information systems meet the intent of Regulatory Guide 1.97, Rev. 3, and ANSI /ANS-4.5, this issue is resolved for the System 80+ Standard Design.
Amendment F A-155                      December 15, 1989
 
s CESSAR !!nhn #
I REFERENCES 1.-  NUREG-0933,  "A Status Report on Unresolved Safety -Issues",
U.S. Nuclear Regulatory Commission, April 1989.
: 2. Regulatoty Guide            1.97, Revision 1, " Instrumentation for Light-Water-cooled ~ Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident",
August 1977.
: 3. ANSI /ANS 4.5, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors", December 1980.
: 4. NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, October 1980.
: 5. Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", May 1983.
Amendment F A-156              December 15, 1989
 
1
: CESSAR !!L"ico..                                                                            :
l II.G.it  POWER SUPPLIES FOR PRE 88URIZER RELIEF VALVES, BLOCK VALVES. AND LEVEL INDICATIONS L
ISSUE Generic Safety Issue (GSI) II.G.1 in NUREG-0737 (Reference 1),
addresses the reliability of the emergency power source which is used for the pressurizer relief (PORVs) and block valves and for the pressurizer level indication in the event of loss-of-offsite power.
The TMI accident demonstrated the need for reliable pressuricer equipment    (e.g.,  the ability. to open or close the PORVs as necessary). Moreover, power supplies used to provide power for the pressurizar PORVs, block valves and level indication may not have    been_ qualified    to-      present    stringent          post-accident sequirements. Several concerns were identified regarding the a'dequacy of the power supplies for the pressurizer equipment particularly with respect to the less-of-offsite power event (see                          ,
identifies new guidance to assure NUREG-0737). NUREG-0737 adequate power for the pressurizer equipment consistent with the                            -
requirements of 10 CFR 50 Appendix A,          (Reference 2) (GDCs 10, 14, 15,~17 and 20).
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.G.1 is that,  in- the event of a          loss-of-offsite power, the above pressurizer equipment shall be supplied with uninterrupted electrical power consistent with the guidance given in NUREG-0737                            +
and_the requirements of 10 CFR 50 Appendix A (GDCs 10, 14, 15, 17 and 20).
i
                                                                                                      ^
RESOLUTION The System 80+ Standard Design incorporates pressurizer equipment                            ,
that is different from current operating plant designs.                          For example,.the Safety Depressurization System (SDS) performs rapid venting and depressurization of the Reactor Coolant System (RCS) when the Auxiliary Spray System is not available (see CESSAR-DC, Section 6.7.1.1 for a description of the SDS).                              Reliable pressurizer level indication is provided in the Nuplex 80+
Advanced Control Complex consistent with the guidance given in NUREG-0737.
The System 80+ Standard Design uses the Post Accident Monitoring Instrumentation (PAMI) and the SDS to monitor and mitigate a variety of beyond design basis events (see CESSAR-DC, Sections 6.7 and 7.5).
Amendment F A-157                          December 15, 1989
 
1 j
CESSAR ;!ninem l
l The equipment may be used during postulated accidents (or during beyond design basis events) to perform a rapid depressurization of the RCS or- to perform feed-and-bleed operations. Since this equipment is designated " safety-related", the systems and components including the pressurizer level indication and the                      l l
safety depressurization valves are qualified to meet expected post-accident conditions.                                                        3 In accordance with the safety-related              1 design requirements, the valves can be supplied from an emergency onsite power source in the event of a loss-of-offsite power. The pressurizer fluid level indication and SDS instrumentation are
  'part of the PAMI and are identified (see CESSAR-DC, Section 7.5, Table 7.5-3). The sensors and displays are capable of operating independently of offsite power for PAMI (see CESSAR-DC, Section 7.5.2.5).
In summary, the safety-related SDS and PAMI are powered from emergency onsite power and can sustain a total loss of offsite power (among other design basis events) and remain functional.
Therefore, since the intent of the guidance given in NUREG-0737 has been met and thus the-requirements of 10 CFR 50 Appendix A, this issue is resolved for the System 80+ Standard Design.
REFERENCES
: 1. NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, October 1980,
: 2. 10    CFR 50- Appendix  A,  " General Design              Criteria  for Commercial Nuclear Power Plants", Office of the Federal Register, National Archives and Records Administration, l
Amendment F A-158                            December 15, 1989
 
T i
P y3_, k' no              i CESSAR:lilW,..ri:<                                                                        ~(Shoot 2 or 5)
EFFECTIVE PAGE LISTING 4
APPENDII B l
Table of contents t
23g3                                                                                hatendment
                    ' i!                                                                                          F
                      - 11                                                                                        F                              ,
f                111-                                                                                      F iv                                                                                        F v                                                                                          F vi                                                                                          F W                      vii                                                                                        F                                ~
l'                    viii                                                                                        F=
ix-                                                                                        F                                3 x"                                                                                        F Xi-                                                                                        F l                      Xii                                                                                        F L                      xiii                                                                                      F                                ,
xiv-                                                                                      F
                      - xv                                                                                        .F-                              ,
ISEI Rana                                                                                Amendment                                7 i
                      - B-1                                                                                        F ll                      B-2'                                                                                      F
!                      B-3                                                                                        F                                .
: B-4                                                                                        F l-                -
F B-5'                                                                                                                        ,
                    - B-6                                                                                        F
                      ' B-7                                                                                        F B-8.                                                                                      F F
B-9'
                      - B-10                                                                                      F L                    ~B-11                                                                                        F l                      B-12                                                                                      F
                      -B-13                                                                                        F l-                      B-14                                                                                      F o                      B-15                                                                                      F i,
B                                                                                      F B-17                                                                                      F c                      B-18                                                                                      F B-19                                                                                      F B-20                                                                                      F L-                                                                                                                Amendment F December 15, 1989
 
                          .                                                                              1 CESSAR EWico.                                                          (Sheet 2 of 5)'
i 1
l 3.
4 EFFECTIVE PAGE LISTING (Cont'd)'
                                                            &PPENDII B                                      ;
,                                                                                                          1
                                                                                                          )
1331 (Cont'd)                                                                        -l EnEB                                                                .-  -
                                                                                        .t :              !
I
'.        . B-21'                                                                  F                      :
l-          B-22                                                                    F B-23                                                                    F
(-    ;
B-24                                                                    F B-25                                                                    F l            B-25                                                                  F l          B-27                                                                    F j          ,B-28                                                                  F L            B-29                                                                  F                    t B-30L                                                                  F                    -'
B-31                                                                    F
          'B-32                                                                    F B-33'                                                                  F                    :
B-34                                                                  F B-35                                                            '
F F
B-36 B-37                                                                  F B-38                                                                  F B-39                                                                  F B-40                                                                  F B-41                                                                  F B-42                                                                  F B-43                                                                  F l'          B-44                                                                  F B-45.                                                                  F                    I B-46                                                                  F B-47                                                                  F B-48                                                                  F B-49                                                                  F l            B-50                                                                  F l'          B-51                                                                  F B-52                                                                  F L-          B-53                                                                  F l            B-54                                                                  F                      -
l            B-55                                                                  F F
B-56
,          -B-57                                                                    F B-58                                                                  F                      ,
B-59                                                                  F B-60                                                                  F B-61                                                                  F B-62                                                                  F Amendment F December 15, 1989 l-
 
  .-k' CESSAR !!nincu..,                                                                                                                            (Shost 3 of S)
F EFFECTIVE PAGE LISTING (Cont'd)-                                                        ,
APPENDII B 1
ISKt (cont'd);
2 age                                                                                                                      amendment B-63                                                                                                                                        F.
B-64                                                                                                                                        F                                  f B-65                                                                                                                                        F B-66                                                                                                                                        F                                  '
B-67                                                                                                                                        F E            - B                                                                                                                                        F B-69                                                                                                                                        F B-70'                                                                                                                                      F B                                                                                                                                      F B-72                                                                                                                                        F
              ~ B-73                                                                                                                                          F
!                B                                                                                                                                      F L-                B-75                                                                                                                                        F'                                ,
l                B-76                                                                                                                                        F B-77                                                                                                                                        F B-78                                                                                                                                      'F                                  <
B-79                                                                                                                                        F B                                                                                                                                      F B-81                                                                                                                                        F
              - B-82                                                                                                                                          F B-83                                                                                                                                        F E
B-84                                                                                                                                        F
~
B-85                                                                                                                                        F                                  i B-86                                                                                                                                        F.
o                B-87                                                                                                                                        F B-88                                                                                                                                        F B-89                                                                                                                                        F B-90                                                                                                                                        F B-91'                                                                                                                                        F B-92                                                                                                                                        F B-93                                                                                                                                        F                                '
B-94                                                                                                                                        F
: B-95                                                                                                                                          F                                >
B-96                                                                                                                                        F B-97                                                                                                                                        F B-98                                                                                                                                        F B-99                                                                                                                                        F B-100                                                                                                                                        F B-101                                                                                                                                        F B-102                                                                                                                                        F                                    l B-103                                                                                                                                        F                                    i B-104                                                                                                                                        F l
Amendment F December 15, 1989 i
 
6 TCESSAR !!!a"icario,i                                                      (Sheet 4 of'5)-
    +
5FFECTIVE PAGE LISTING (Cont'd)                                                      ,
t APPENDII B 23X1 (Cont'd)-
Zaga                                                          amendment B-105                                                                    F
        - B-106-                                                                  F
        - B-107                                                                    F B-108                                                                    F                            s B-109                                                                    F B-110                                                                    F B-111                                                                    F
        'B-112                                                                    F B-113                                                                    F' B-114                                                                    F B-115                                                                    F B-116                                                                    F B-117              ,                                                    F B-118                                                                    F B-119                                                                    F B-120                                                                    F' r      'B-121                                                                    F                            ,
l-;      B-122                                                                    F Tables                                                  Amendment Bl.3.1                                                                  F B3.1.1-1                                                                F B3.1.2-1                                                                F B3.1.2-2~                                                                F B3.1'.3-1                                                                F B3.1~.4-1                                                                F B3.1.5-1 (Sheet 1)                                                      F L-        B3.1.5-1 (Sheet 2)                                                      F B3.1.6-1                                                                F B3.1.7-1                                                                F B3.1.8-1                                                                F B3.1.9-1                                                                F B3.1.10-1                                                                F
;.        B3.1.11-1                                                                F B3.1.12-1 (Sheet 1)                                                      F B3.1.12-1 (Sheet 2)                                                      F B3.1.13-1                                                                F B3.2-1                                                                  F B3.2-2 (Sheet 1)                                                        F
;        B3.2-2 (Sheet 2)                                                        F Amendment F December 15, 1989
 
s- %
e            >
f                1CES.SARTHF#,c.n ..                                          (Sh2$t 5 of 5)-                                        !
L EFFECTIVE PAGE LISTING (Cont'd)
APPENDII B i
Tables (Cont'd)                        Amendment
                  - B3 '. 2 -2      (Sheet 3).                                F
                  .B3.2-21 (Sheet 4).                                          F B3.2-2 (Sheet-5)                                          F Figures                                Amendment B2.0-1                                                    F B2.3-1                                                    F B3.1;1-1                                                  F
!'                    B3.1.2-1~                                                F B3.1.3-1                                                  F B3.1'.4-1                                                F B3.1.5-1                                                  F
:B3.1;6-1                                                    F B3;1.7-1~~
F B3.1.8-1                                                  F B3.1.8-2                                                  F,                                                    ,
i                      B3.1.9-1                                                  F l
l~
B3.1.10-1                                                F l                  'B3.1.11-1'                                                  F
                  .B3.1.12-1                                                    F B3.1.13-1                                                F B3 .1'.13 -2                                              F I
N                                                                                                                                        l l
I'                                                                                                                                      l 1
Amendment F                                              1 December 15, 1989
 
                                  ,      -.                                  ~                      .                    .  . . . .-                .      ~.-.        .  . .            - _. _ . _..    . . . - . ..          .
                          )                                                                                                                                                                                                              '
                                                                +:
ge. hI        ,          ,:g g. , -                        .
j
      >            g        ,-3n-                                                                  >
                            .Ni. !
8 g Dell *N'                            -
                                                                                                                                                                                                                                              ]
                    "y N.                  ereMsu CERTIFICATl!N '                                                                                                                                                        ,
i
            ;                                                                                                                                                                                                                            n
                      ,1                                                                                                                                                                                                                    .,
                                                                                                                                                                                                                                        ' I
                                                                                                                                                                                                                                        ^
t l
t 1
I
    . s r                  .
1
                            .1 J
            ;                                                                                                                                                                                                                                  l
                                    .r 1
        ?
1 I
                                                                                                                  . APPENDII.B E                                                                                                PROBABILISTIC RISK ASSESSMENT 1                                                                                                                          FOR THE g-                                                                  SYSTEM 80+TM Ll..
i                      .                                                                                    STANDARD PLANT DESIGN s
I g,
1 4
t:
i-(.5 1i
                                                                                                                                                                                                                                            +
l :.
ll f
j .,                                                                                                                                                                                                                                    -i 1,
f.
i      k ';
i
(-
I Amendment F December 15, 1989 0
            ;,                    n.-,        , . . . - - - ,      , , - . .  .--,.,,-...~.--,n.-              ,--,e....              v., ,, , , , ,-,.  ~,v,-re. , - - ,,,    ,,,-,---r,-              ,a      ,-.-a. . a--. --
 
yw                            a    ,.      ,                                                  ,,                          ,
  !          ?        ,                                                                                                                        ,
I b                            '
                                  !CESSARiinkm,                                                                                          l
        ~,g.
l TABLE OP CONTENTS                                              1 4.
APPENDIE B                                                l l .-
Section                Subient                                                    Pace No.
: 1. 0 -                  INTRODUCTION                                            - B-l' .
1.1                    PURPOSE                                                  'B-1                      I
                                  -1.2                    SCOPE                                                    B-1                  .)
 
==SUMMARY==
OF RESULTS                                        B-1
                                  .1.3-2.0.                    METHODOIDGY                                              'B-3                    ,
a                                2 .' 1                PLANT FAMILIARIZATION                                    B-5' n
L                                  2.2                    ACCIDENT'SEOUENCE DEFINITION                              B-5 9
B-8                - +
2.3                    SYSTEM MODELING 2.4                    DATA ASSESSMENT                                          .B-11                .,
2.5                    ACCIDENT SEOUENCE OUANTIFICATION                          B-12L                  ,
2.6'                  EXTERNAL EVENTS ANALYSIS                                  B-14                g 2.7                    SQt[I&INMENT RESPONSE ANALYSIS                            P-15 2.D.                  CONSEQUENCE ANALYSIS                                      B-25                  q L
* U                                '2.9                    ANALYSIS GROUNDRULES-                                      B-16 l
L                                  2.10                ' DESCRIPTION OF COMPUTER CODES                              B-17                -)
                                                                                                                                          . ;l 2.10.1:                IRRAS                                                      B-17                a B                                  '2.10.2                CESAM                                                      B-17
!                                                                                                                                              1 1
2.10.3-                CENTS                                                      B-17                      l 1:
2.10.4                MAAP-DOE                                                  B-18 l-p                                  2.10.5                CRAC2                                                      B-19 l
l Amendment F                          !
L                                                                                        i                December 15, 1989 L
          '-( '    . .
 
            .    -      . .      ..    .-.  . . . -      - ~ . - _ .      . - . -
l l
3115!iAllt.!Entricar...                                                              !l l
1 l
l TABLE OF CONTENTS (Cont'd)                                        !
i APPENDIX B                                                I 1
l Section        Subiect                                              Page No.
3.0            ACCIDENT SEQUENCE DETERMINATION                      B-21 1
3.1            EVENT TREE ANALYSIS                                  B-21 3.1.1          LARGE LOCA                                          B-21            '
3.1.1.1              Event Tree 1 Elements                          B-21 3.1.1.1.1            Large LOCA Initiators                          B-21 3.1.1.1.2            Safety Injection Tank Injection                B-22
;  3.1.1.1.3            Safety Injection. System Injection              B-22~
p  3.1.1.1.4            Containment Spray Cooling                      B-23 3.1.1.2              Maior Decendencies.                            B-24 3.1.1.3              Maior Recovery Actions                          B-24 3.1.1.4              Core Damace Secuence Ouantification            B-24
  . 3.1.2          MEDIUM LOCA                                          B-25 3.1.2.1              Event Tree 2 Elements                          B-26 3.1.2.1.1            Medium LOCA Initiators                          B-26 3.1.2.1.2            Safety Injection System Injection              B-26 l  3.1.2.1.3            Containment Spray Cooling                      B-27 l
l 3.1.2.2              Maior Decendencies                              B-27 3.1.2..*            Maior Recovery Actions                          B-27
(  3.1.2.4              Core Damaae Secuence Ouantification            B-28 l                                                                                          l l
3.1.3          SMALL LOCA                                          B-29 3.1.3.1              Event Tree 3 Elements                          B-29 3.1.3.1.1            Small LOCA Initiators                          B-29 i  3.1.3.1.2            Safety Injection System Injection              B-29 L  3.1.3.1.3            Aggressive Secondary Cooldown                  B-30 l  3.1.3.1.4            Shutdown Cooling System Injection              B-30              l 3.1.3.1.5            Deliver Feedwater                              B-31              )
l 3.1.3.1.6            Long-Term Decay Heat Removal                    B-31              l 3.1.3.1.7            Safety Depressurization (Bleed)                B-32              )
Amendment F 11                December 15, 1989 i
                                                                                      -.I
 
i    e
:CESSAR1Ene m.,
TABLE OF CONTENTS (Cont'd)
      -                            APPENDIX B                                            l Section    Subiect                                                Pace _No1 q
3.1.3.1.8-      Cooling the In-containment Refueling              B-32 Water Storage Tank                                              I 3.1.3.2          Maior Denendencies                                B-33 3.1.3.3        .Maior Recovery Actions                            B-33        l 3.1.3.4          core Damace Secuence Ouantification              B-34 3.1.4      STEAM GENERATOR TUBE RUPTURE                            B-35 3.1.4.1          Normal Transient Proaression                      B-35      2 3.1.4.2          Accident Proaression with coincident              B-37        !
LD.QE 3.1.4.3          Event Tree 4 Elements                            B-37 3.1.4.3.1        Steam Generator Tube Rupture                      B-37 Initiators l                          Safety Injection System Injection                B-38 1          3.1.4.3.2 3.1.4.3.3        Aggressive Secondary Cooldown                    B-38 3.1.4.3.4        Shutdown Cooling System Injection                B-39 3.1.4.3'.5      Deliver Feedwater                                B-39 3.1.4.3.6        RCS Pressure Control                              B-40 3.1.4.3.7        Long-Term Decay Heat Removal                      B-40 3.1.4.3.8        Unisolable Leak in Ruptured Steam                B-41 Generator 3.1.4.3.9        Maintain Secondary Heat Removal (MSHR)            B-42 3.1.4.3.10      Safety Depressurization (Bleed)                  B-42 3.1.4.3.11      Safety Injection (Feed)                          B-43 3.1.4.3.12      Cooling the In-containment Refueling              B-43 Water Storage Tank 3.1.4.4          Maior Denendencies                                B-44 3.1.4.5          Maior Recoverv Actions                            B-44 3.1.4.6          Core Damaae Secuence Ouantification              B-45 l
Amendment F iii                        December 15, 1989
 
w
                    ,I 5!iJLit.s!.rificar..:
1 TABLE OF CONTENTS (Cont'd)
APPENDII B Section                  Subiect                                                              Pace No.
3.1.5                    LARGE SECONDARY SIDE BREAKS                                          B                    3;1.5.1                      Event Tree 5 Elements                                          B-48                ;
                - 3.1.5.1.1                      Large Secondary Side Break Initiators                          B-48 3.1.5.1.2                    Safety. Injection System Injection                              B-48 3.1.5.1.3                    Deliver Emergency Feedwater-                                    B-48                ;
3.1.5.1.4                    Long-Term Decay Heat Removal                                    B-49                  1 3.1.5.1.5                    Safety Depressurization (Bleed)                                B-50 3.1.5.1.6                    Safety Injection (Feed)                                        B                    3.1.5.1.7                    Cooling the In-containment Refueling-                          B-50 Water Storage Tank                                                                  u i
3.1.5.2                      Maior Decendencies                                              B-51                  .
l 3.1.5.3                      Maior Recoverv Actions                                          B-51 3.1.5.4                      Core Damaae Secuence cuantification                            B-51 l                  3.1.6                    TRANSIENTS - LOSS OF FEEDWATER                                      B-53                  i 4
3.1.6.1                      Event Tree 6 Elements                                          B-53                  l 3.1.6.1.1                    Transient Initiators.                                          B-53                  i 3.1.6.1.2                    Deliver Emergency Feedwater                                    B-54 3.1.6.1.3                    Long-Term Decay Heat Removal                                    B-54                  l 3.1.6.1.4                    Safety Depressurization (Bleed)                                B-55                  l 3.1.6.1.5                    Safety Injection (Feed)                                        B-55                  l 3.1.6.1.6                    Cooling the In-containment Refueling                            B-55 Water Storage Tank
                                                                                                                ~
3.1.6.2                      Maior Decendencies                                              B-56              -
3.1.6.3                      Maior Recovery Actions                                          B-56                  l 3.1.6.4                      Core Damaae Secuence Ouantification                            B-57 l                                                                                                                                        l 3.1.7                    OTHER TRANSIENTS                                                    B-59                  l l                  3.1.7.1                      Event Tree 7 Elements                                          B-59 3.1.7.1.1                    Transient Initiators                                            B-59 l                  3.1.7.1.2                    Deliver Feedwater Flow                                          B-59 i                  3.1.7.1.3                    Long-Term Decay Heat Removal                                    B-60 Amendment F iv                                December 15, 1989
 
CESSAR n#,c.m TABLE OF CONTENTS (Cont'd)
APPENDIX B Section                                      Subiect                                                                          Pace No.
3.1.7.1.4                                          Safety Depressurization (Bleed)                                            B-61 3.1.7.1.5                                          Safety Injection (Feed)                                                    B-61
                  . 3.1.7.1.6                                          Cooling the In-containment Refueling                                      B-61 Water Storage Tank.
3.1.7.2                                            Maior Denendencies                                                        B-62 3.1.7.3                                            Maior Recoverv Actions                                                    B-62 3.1.7.4                                            Core Damaae Secuence Ouantification                                        B-63 3.1.8                                        LOSS OF OFFSITE' POWER AND STATION BLACKOUT                                      B-65          '
3.1.8.1                                            Normal Event Proaression                                                  B-65 3.1.8.2                                            Loss of Offsite Power Event Tree                                          B-66 3.1.8.2.1                                          Event Tree 8 Elements                                                      B-66      ,
i'                    3.1.8.2.1.1                                              Loss of,0ffsite Power Initiators                                    B-66 3.1.E.2.1.2'                                            Primary Safety Valve Reseat                                          B-67 L                      3.1'.8.2.1.3                                            Safety Injection System Injection                                    B-67 l-                    3.1.8.2.1.4                                              Deliver Erergency Feedwater                                          B-67 l                    3.1.8.2.1.5                                              Long-Term Decay' Heat Removal                                        B-68 l                    3.1.8.2.1.6                                              Safety Depressurization (Bleed)                                      B-69 L                    3.1.8.2.1.7                                              Safety Injection (Feed)                                              B-69 l                    3.1.8.2.1.8                                              Cooling the In-containment                                          B-69
!                                                                              Refueling Water Storage Tank 3.1.8.3                                            Maior Decendencies                                                        B-70 3.1.8.4                                            Maior Recovery Actions                                                    B-70 3.1.8.5                                            Core Damaae Secuence Ouantification                                        B-71        '
l, 3.1.8.6                                            station Blackout Event Proaression                                        B-71 3.1.8.6.1                                          Station Blackout Fault Tree Elements                                      B-72 3.1.8.6.2                                          Major Dependencies                                                        B-72 3.1.8.6.3                                          Major Recovery Actions                                                    B-72 Amendment F v                                                  December 15, 1989
 
1    1
:                                                                                                      1
    .n.
CESSARHMiem..
                                                                                                                            )
i    G ;-
1; TABLE OF CONTENTS (Cont'd)
APPENDII B.
1 Bootion          subient                                            Page No.                          l j
3.1.9'            LOSS OF' COMPONENT COOLING' WATER                  B-73 li                                                                                                                        'i 3.1.9.1                Event Tree 9 Elements                        B                            I L                    ' 3.1.9.1.1              Transient Initiators                          B-73 3'.1.9.'1.2            Deliver Feedwater                            B-74                            i 3.1.9.1.3;              Long-Term Decay Heat Removal                  B-74                              1 3.1.9.1.4              Safety'Depressurization (Bleed)              B-75 3.1.9.1.5              Safety Injection (Feed)                    .
B-75                            ,
3.1.9.1.6              Cooling the In-containment Refueling          B-76                              I Water Storage Tank 3.1.9.2-                Maior Denendencies                            B-76 l
                                                                                                                            ~
                    . 3 .1. 9. 3 :          Maior Recovery Actions                        B-77
                    - 3.1.9.4.              ' core Damace Secuence Ouantification          B-77 l
                    - 3.'1.10          LOSS OF 125 VDC VITAL BUS                          B-79                            ;
l-                    3.1'.10.1-            . Event Tree 10 Elements                      B-79 L                                                                                                                            I 3.1~10.1.1
                            .                Transient Initiators                          B-79 3.1.10.1.2              Deliver Feedwater                            B-79
          ,          3.1.10.1.3              Long-Term Decay Heat Removal                  B-80
                    . 3.1.10.1.4            Safety Depressurization.(Bleed)              B-81 3.1.10.1.5              Safety Injection (Feed)                      B-81 3.1.10.1.6              Cooling the In-containment Refueling          B-81 Water Storage Tank                                                            j 3.1.10.2                Maior Decendencies                            B-82 3.1'.10.3              Maior Recoverv Actions                        B-82                              1 I
3.1.10.4                core Damace Secuence ouantification          B-83                              i l
3.1.11            LOSS OF 4.16 KV VITAL BUS                          B-85                            I 3.1.11.1                Event Tree 11 Elements                        B-85 3.1.11.1.1              Transient Initiators                          B-85 3.1.11.1.2              Deliver Feedwater                            B-85 3.1.11.1.3              Long-Term Decay Heat Removal                  B-86 Amendment F p
vi                December 15, 1989 l
l
                                                                                                              . . _ , . , ,l
 
E    . v; 2
8 CESSAR E!ntricario.                                                                                                                  l l
V i
l TABLE OF CONTENTS (Cont'd)
APPENDIX B Section                Subiect                                                                                    Pace No.
3.1.11.1.4                        Safety Depressurization (Bleed)                                                  B-87 y        3.1.11.1.5                        Safety Injection (Feed)                                                          B-87              l 3.1.11.1.6                        Cooling ti.) In-containment Refueling                                            B-87              l Water Storage Tank                                                                                  ,
                                                                                                                                                .1 3.1.11.2                          Maior Deeendencies                                                                B-88            i 3.1.11.3                          Maior Recovery Actions                                                            B-88 3.1.11.4                          core Damaae Secuence Ouantification                                              B-89              1 3.1.12                  ANTICIPATED TRANSIENTS WITHOUT SCRAM                                                      B-91 3.1.12.1                          ATWS Descriotion                                                                B-91 3.1.12.2                          ATWS Event Tree Elements                                                        B-93 3.1.12.2.1                        ATWS Initiators                                                                  B-93 3.7. 12.2.2                        Adverse Moderator Tamperature                                                    B-94 Coefficient (MTC) 3.1.12.2.3                        Primary Safety Valve (PSV) Stuck Open                                            B-94          ~
l 3.1.12.2.4                        Consequential Steam Generator Tube                                              B-94 p                                              Rupture (SGTR)'
l          3.1.12.2.5-                        Deliver Emergency Feedwater                                                      B-94 L          3.1.12.2.6                        Deliver Boron Via Charging Pump                                                  B-95 L          3.1.12.2.7                        Safety Depressurization                                                          B-95 3.1.12.2.8                        Safety Injection                                                                B-95 3.1.12.2.9                        RCF Pressure Control                                                            B-96 3.1.12.2.10                        Long-Term' Decay Heat Removal                                                    B-96              l 3.1.12.2.11                        Unisolable Leak in the Ruptured                                                  B-97              J Generator                                                                                          ,
3.1.12.2.12                        Maintain Secondary Heat Removal (MSHR)                                          B-98              !'
3.1.12.2.13                        Cooling the In-containment Refueling                                            B-98 Water Storage Tank 1
3.1.12.3                          Maior Decendencies                                                              B-99 3.1.12.4                          Maior Recovery Actions                                                          B-100 3.1.12.5                            Core Damaae Secuence Ouantification                                            B-100 l
Amendment F vil                                December 15, 1989
            +- a e ,- - -a w- - --,-.    ,,wgwa  .r- - - -  yo wr-,,ww- y ww-a w wwwewy,--w-wvp y-q-yrw-    re w++ w - --+gw-  9
                                                                                                                                    - =p- --
 
    \
C E S S A R iinine m a                                                                                                                        ,
6
                                                                                                                                                          ,e TABLE OF CONTENTS (Cont'd)                                                                              ,
l:                                                                                                                                                          1 APPENDIX ~B l                                                                                                                                                          i Section                        Subiect                                                                                    Pace No.
3.1.13                          INTERFACING SYSTEM LOCAS-                                                                B-101
(            3.1-.13.1-                                Interfacina System LOCA Via SCS                                                  B-102 Return Lines 3.1.13.2                                  Interfacina System LOCA Via SCS                                                  B-106 l:
,                                                      Suction Lines I-          3.1.13.3                                  Interfacina System LOCA Frecuency                                                B-110 3.1'.14                          VESSEL RUPTURE                                                                            B-113 l-'
b            3.2                              $UMMARY OF CORE DAMAGE"FREOUENCY                                                          B-115 FOR INTERNAL EVENTS
: l.          4.O                              EXTERNAL EVENTS EVALUATION                                                              B-117 I --
5.0                              CONTAINMENT RESPONSE ANALYSIS                                                            B-117              ;
          "r 6.O                              CONSEOUENCE ANALYSIS                                                                    B-117 7.0                             
 
==SUMMARY==
AND CONCLUSIONS                                                                  B-117
 
==8.0                              REFERENCES==
FOR APPENDIX B                                                                B-118 E.
L 1
i l
I e
i
!                                                                                                                        Amendment F viii                                        December 15, 1989 E .. - . _            _ . . - . _ - - _ . . .                  . _ _ _ _ . _ . _ . . - _ - _ . . . _ .    - . - _  _ . . _ _ . _ . . _ . _ , . . .
 
7
:CESSAR1!nL m                                                                                                                                          J s                                                                                                                                                                  u L
1
.                                                                                                                                                                    i LIST OP' TABLES 1
APPENDIX B g      -
Table                  subiect w"
Bl.3              Comparison of Severe Accident-Frequencies                                                                                      d 1
I' B3.1.1-1              Core Damage Frequency Contributions for Large Loss of Coolant Accident (LLOCA) Core Damage Sequences B3.1.2-1              Core Damage Frequency Contributions for Medium Loss of Coolant Accident 1 (MLOCA1) Core Damage Sequences B3.1.2-2              Core Damage Frequency Contributions for Medium Loss of Coolant Accident 2- (MLOCA2) Core Damage Sequences
                                                                                                                                                        . .        J
            ,B3.1.3-1              Core Damage Frequency Contributions for Small Loss                                                                    '
of Coolant Accident (SLOCA) Core Damage Sequences B3.1.4-1              Core Damage Frequency Contributions for Steam Generator Tube Rupture (SGTR) Core Damage. Sequences B3.1~.5-1              Core Damage Frequency Contributions for.Large Secondary Side Break (LSSB) Core Damage Sequences B3.1.6-1              Core Damage Frequency Contributions for Loss of
                                    - Feedwater Flow (LOFW) Core Damage Sequences B3.1.7-1              Core Damage Frequency Contributions for Other                                                                                  J Transients (TOTH) Core-Damage Sequences
                                                                                                                                                                    -(
B3.1.8-1                Core Damage Frequency Contributions for Loss of                                                                                l Offsite Power (LOOP) Core Damage Sequences                                                                                      I l
B3.1.9-1                Core Damage Frequency Contributions for Loss of Component Cooling Water Div 2(B) (CCWB) Core Damage Sequences B3.1.10-1                Core Damage Frequency Contributions for Loss of                                                                                I 125 VDC Vital Bus B (125VB) Core Damage Sequences                                                                            q B3.1.11-1              Core Damage Frequency Contributions for Loss of                                                                                l 4.16 KV Vital Bus B (410KB) Core Damage Sequences l
l Amendment F ix                                          December 15, 1989
              ---a-  -- -  -----  _  -n__ - e, ----.-._--e  w  - - - , ,  -a n.--.,,,r--we ,--,we,aew,,,,are      . , - - - . - e ,,.,---,~ea  , ,    ----ei
 
E CESSAR !!!Wicms.
l l
LIST OF TABLES (Cont'd) 1.
APPENDIX 3 l-                  Table            8ubiect B3.1.12-1        Core Damage Frequency Contributions for Anticipated Transient Without Scram (ATWS) Core Damage Sequences l'~
L                    B3.1.13-1        Potential: Paths for Interfacing System LOCAs B3.2-1          Core Damage Frequency Contribution by Initiating Event B3.2-2          Component Importances For System 80+ PRA l
l
[  . . ..
          .t-5 l
l 2
l Amendment F x                    December 15, 1989
 
            \
t  a p
CESSAR1!! h on-LIST-OF FIGURE 8' g
APPENDIE B Figure        subient B2.0-l'        Major PRA~ Tasks B2.3.1:        Fault Tree.'Symbology B3 '.1.1-1    Large LOCA Event Tree B3.1.2    ~ Medium 1LOCA' Event Tree L                    B3.1.3      Small LOCA Event-Tree B3.1'. 4 -Steam Generator Tube Rupture Event Tree B3.1.5-1      Large Secondary Side Break Event Tree B3.1.6-1      Loss Of Main Feedwater Event Tree B3.1'.7-1      Other Transients Event Tree B3.1.8-1      Loss of offsite: Power Event Tree i                B3.1.8-2      Station Blackout with Battery Depletion Fault Tree B3.1.9-1      Loss of Component Cooling Water Event Tree B3.1.10-1      Loss of 125 VDC Vital Bus Event Tree B3.1.11-1      Loss of 4.16-KV Bus Event Tree B3.1.12-1      ATWS Event Tree B3.1.13-1      Check Valve Arrangement in Shutdown Cooling System Return Lines B3.1.13-2    Motor Operated Valves Arrangement in Shutdown Cooling System Suction Lines 1
Amendment F xi                Decetaber 15, 1989
 
i l          LCESSARinsnen..
L-              .o                                                                            i t
r i                                      LIST OF ACRONYMS                                      1 APPENDIE B ac:      alternating current ADV      Atmospheric Dump Valve                                              1 ALWR    Advanced Light Water Reactor E              -ANS    'American Nuclear Society                                                ,
L                ANSI    American. National Standards Institute                                l AO      Auxiliary Operator ADO      Anticipated Operational Occurrence ARO      All Rods out                                                          i 1 ARTS    Alternate Reactor Trip System                                          l ARSAP    Advanced Reactor Severe Accident Program
                'ATWS      Anticipated' Transient Without Scram BNL      Brookhaven National Laboratory BOP-    Balance of Plant BWR. Boiling Water Reactor
                .C-E;  . Combustion Engineering CCS. Component Control System CCW      Component Cooling Water .    ..
                'CCWS      Component Cooling Wdter System CDC-    Control. Data Corporation                                              I CEA      Control Element Assembly                                            Y CEDM:    Control Element Drive Mechanism                                        l CENTS. Combustion Engineering Nuclear Transient Simulator                    l CEOG. Combustion Engineering Owners Group CET      Containment Event Tree CIAS    Containment Isolation Actuation Signal CPC      Core Protection Calculator CRAC    Calculation of Reactor Accident Consequences
                .CSAS      Containment Spray Actuation Signal CS.      Containment Spray CSET    Containment Safegaurds Event Tree CSS      Containment Spray System CST      Condensate Storage Tank DEC    ' Digital Equipment Corporation DG      Diesel Generator DOE      Department of Energy DVI      Direct Vessel Injection ECCS    Emergency Core Cooling System                                          l l                EDS      Electrical Distribution System
                'EFAS    Emergency Feedwater Actuation Signal EFW      Emergency Feedwater                                                    i EFWS    Emergency Feedwater System                                            l EPRI    Electric Power Research Institute                                    j ERF      Error Factor ESF      Engineered Safety Features                                            ,
ESFAS    Engineered Safety Features Actuation Signal (or System)
FP      Fission Products                                                      .
l Amendment F xii                    December 15, 1989 l
 
ll CESSAR !nWne.no.
LIST OP ACRONYMS'(Cont'd)-
APPENDIX 3 g-            unit of acceleration equal to.the acceleration of gravity gpm          Gallons Per Minute HCR            Human Cognitive Reliability HEP-          Human Error Probability HPSI          High Pressure Safety Anjection IBM            International' Business Machines, Inc.
IDCOR          Industry Degraded Core Rulemaking Program-
,'                    Individual Plant Evaluation IPE IPEM          Individual Plant Evaluation Methodology L
IREP          Interim Reliability Evaluation Program L      IRRAS          Integrated Reliability and Risk Assessment System                                                                                    1 IRWST          In-containment Refueling Water Storage Tank KAG            Key Assumptions and Groundrules KV            Kilovolts                                                                                                                  3 LOCA-          Losslof Coolant Accident                                                                                                    .;
LOOP          Loss of Offsite Power LSSB          Large Secondary Side Break LWR            Light Water-Reactor MAAP          Modular Accident Analysis Program MCC            Motor Control Center MFIV          Main Feedwater Isolation Valve                                                                                              -;
MFW            Main Feedwater System                                                                                                        _
MLD            Master Logic Diagram                                                                                                        ;
mph            miles per hour                                                                                                                ,
MSIS          Main Steam-Isolation Signal MSIV          Main Steam Isolation Valve                                                                                                            i MSSV          Main Steam Safety Valve NRC            Nuclear Regulatory Commission NREP          National Reliability Evaluation Program NSSS          Nuclear Steam Supply System
      - PC            Personal Computer PDS          Plant Damage State L        pga          -peak ground acceleration Plant Protection System
[        PPS PRA          Probabilistic Risk-Assessment L
L PSA          Probabilistic Safety Assessment psia          pounds per square inch absolute PSV          Primary Safety Valve PTS            Pressurized Thermal Shock PWR            Pressurized Water Reactor RCGVS        Reactor Coolant Gas Vent System RCP          Reactor Coolant Pump RCS          Reactor Coolant System l        RHR          Residual Heat Removal I
RO            Reactor operator L        RPS          Reactor Protection System l                                                                                                                  Amendment F xiii                                        December 15, 1989
 
c-b;      ..
  ": t.
          ~
b                  C E S S A R Ei! Wicui .
o LIST OF ACRONYMS-(Cont'd) l-D                                                  APPENDIX B RWST-      Refueling Water Storage Tank.
L                  SCS      -Shutdown Cooling System V                  SG-        Steam Generator
: p.                  SGTR      Steam Generator Tube Rupture
!                  SI        Safety Injection R
SIAS      Safety Injection Actuation Signal SIT-      Safety Injection Tank SPS        Supplementary Protection System SRO        Senior Reactor Operator SRP        Standard Review Plan STA        Shift Technical Advisor TBS        Turbine' Bypass System g            VDC        Volts - Direct Current 4
l l      4 l.
l Amendment F xiv                                                                December 15, 1989 l
 
f
        . CESSAR1!!aemo.                                                                                                          1 1.O        INTRODUCTION                                                                                                  l 1.1        PURPOSE One of the        requirements of 10 CFR' Part 52 III is that- an                                                      i application    for. design certification must contain a design                                                          l 1
specific Probabilistic Risk Assessment (PRA).                                                  This Appendix documents the results of the probabilistic risk assessment for the System 80+ Standard. Plant Design.                                                                                !
The System 80+ Standard Design PRA has three primary purposes.
The first purpose is to identify the dominant contributors to                                                        '
severe. accident risk.                  The second purpose is to provide an analytical tool for evaluating the impact of design modifications on core damage probability.and the overall risk to the health and safety of the public.              The final purpose is to calculate the core damage frequency and large release frequency for the System 80+
Standard Design.
1.2        SCOPE                                                                                                    i The System 80+ Probabilistic Risk Assessment is a Level III PRA for the System .80+ Standard Plant Design as described in
        . CESSAR-DC.      This PRA- addresses both internal and external initiators of accident sequences which lead to core damage.                                                            .
          . Bounding plant site characteristics were used for the evaluation of external events such as seismic and tornado strike events and for evaluating public risk.
1.3'       
 
==SUMMARY==
OF RESULTS
!          The System 80+ Standard Design evolved from the System 80 via the incorporation of design enhancements to improve the operation and safety of the plant. The most significant advances from System 80 to System    80+ are in. the design of the engineered safety features. Four-train systems are provided for safety injection, emergency    feedwater                and  integrated                            shutdown cooling        and containment spray.          A        dedicated        safety                    depressurization      system provides feed-and-bleed capability. An alternate AC power source helps cope with loss of power events.                                                In addition, a larger pressurizer and larger steam generators provide a slower, more manageable response to accidents. As shown on Table Bl.3-1, the combined design improvements of System 80+ have decreased the likelihood    of    a    severe accident attributable to                                          internal initiators    by      more            than  two          orders                    of    magnitude    with L
across-the-board improvements for every internal initiator.                                                The design hes progressed to the point that very low probability initiators such as " Vessel Rupture" represent an appreciable fraction of the residual risk.
Amendment F B-1                                        December 15, 1989
 
                                                                              =-- -    - -    - -
                                                                                                        -~                -- -
g . ut, y ;                                          '
id;t - p y                A. . '.                                                                                              t, dj -
        ' -'N {.!
X CESSAR !!Nne.mt                                                                            !
l 0,          .
                  \
t                                                                                                      .
  ,          '!) -                                                                                                                )
l
* 1 7.,.
          ; M ' ,. .'
i        <
i
                                                                                                                                'l l
              't 4
g                                                THIS PAGE INTENTIONALY BLANK I;                                                                                                                                4 l'
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                                                                                                                              .1
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i-                                                                                                                                  1 Amendment F B-2              December 15, 1989
                    +    r                                  e--                            frajp.y, -
                                                                                                            %- gve, we yp.
 
1 CESSAR88% m=                                                                                            !
TABLE B1.3-1                                                      {
COMPARISON OF SEVERE ACCIDENT FREQUENCIES                                                  j severe heeldent Frequency Initiatina Event                              SYSTEM 80+                        SYSTEM 80      i Large I.oss-of-Coolant-Accident                      6.12E-8                            1.57E-6 Medium Loss-Of-Coolant-Accident                      1.16E-7                            3.59E-6 Small Loss-Of-Coolant-Accident                        4.31E-8                            9.41E-6
  .Large Secondary Side Break                            2.74E-10                          9.04E-7        ,
Steam Generator Tube Rupture                          1.38E-7                            1.05E-5        1 Transients                                            2.30E-8                            1.17E-5
{
Loss of Feedwater Flow                                    5.84E-9                  N/S Other Transients                                          4.64E-9                  N/S
!      Loss of Component Cooling                                  1.25E-B                  N/A Water L
Loss of 4.16 Kv Vital Bus                                  2.75E-11                N/S            l Loss of 125 VDC Vital Bus                                  2.61E-12                N/S Loss of Offsite Power (LOOP)                          9.14E-8                          3.78E-5 l  including Station Blackout with                                                                        t
;  Battery Depletion Anticipated Transient Without                          1.97E-7                          4.79E-6        ;
Scram (ATUS)
Interfacing System ICCA                                3.01E-9                          4.48E-9 Vessal Rupture                                          1.00E-7                          1.00E-7 t
TOTAL:          7.73E-7                          8.14E-5 NOTES:    N/A means Not Applicable, N/S means Not Calculated Separately.
Amendment F December 15, 1989 l
 
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I CESSARHMcm.,                                                                              !
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2.0          METHODOLOGY J
This PRA was performed in two phases.            In the first phase, event tree g fault tree models were developed for the System 80 design      . These models were used to estimate the baseline core                      !
damage frequency for a generic plant with a System 30 NSSS and go identify the dominant core damage contributors.                      (System 80 ,
described in CESSAR-F and approv                      the NRC(FDA-2), was the starting point for the System 80+g by PRA.) The second phase was an interactive process in which the system fault tree models developed in phase 1 were modified to reflect proposed system design changes and enhancements.                The revised models were                  ;
evaluated to determine the impact of the design changes on core                            l damage frequency and dominant core damage contributors. This                              ,
information was fed back to the system designers for consideration in the design process.              At the conclusion of the                ,
design process, the PRA evaluation models represented the final System 80+ Standard Plant design. The analysis was then extended to a full level III PRA.
Standard methodologies were used in this analysis.                    The level I        l (core damage frequency) portion of the analysis is equivalent to the baseline probabil tic safety analysis (PSA) described in the PSA Procedures Guide        I and the methodologies employed in this analysis are consistent, within the sco pe and intent of this with methodologies outlined Ln the PSA Procedures analyg,and Guide              methodologies described in the PRA Procedures Guide I4) . The  methodologies used in this PRA comply with the recommendations of the "PRA Key Assumptions and Groundrules" in Appendix (Sp      to    chapter 1      of    the    EPRI    ALWR Requirements              ,
Document      . The  small  event  tree  /large    fault tree approach was used for the evaluation of core camage frequency. The external events analyses were performed by an independent contractor under the Advanced Reactor Severe Accident Program ( ARS AP) .                          The external event evaluation methodology is summarized in Section 2.6, and the results are presented and discussed in Section                        4.
The methodologies used for the level II (source term) wereconsistentwiththemethodologiesusedinNUREGySOglysis                        ,  the methodologies described in the PRA Procedures Guide                      and those methodolges          recommended    in    the  EPRI    ALWR      Requirements Document          The level III analyses (consequences) also used methodologies epistent with those described in the PRA Procedures Guide          and the y thodologies recommended in the EPRI ALWR Requirements Document            . Figure B2.0-1 shows the major tasks in this analysis.          The following sections describe these tasks and associated methodology in greater detail, i
l Amendment F B-3                        December 15, 1989
 
CESSAR tilhn.
I i
l THIS PAGE INTENTIONALLY BLANK-                                                                              ,
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4 Amendment F E-4                                  December 15, 1989
 
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1                                                                      :
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o                                                                                                ;
o                                                                                              .
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                                                                        }            .
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l          e Ji Amendment F December 15, 1989 Figure J g 7g                                            Major PRA Tasks                                                      B2.0 1
 
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                                .ms.                                                                                                                                                                                                                                          ._\s-'t'
:S.? c' ' 3                                      4                                                                                                                          ,r, ..
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                                              .-p
? 'i                                        :      -
                                                                                                    ..                                                          1                        1                                                                                          ,                                                                                                            .4.-
,2            L                              y . l Y.:,, . -( y                                                                                                    .r 1
A L.'
4 Y .-                                                  -I j
                                                                                                                                                              ''' k .{                                        ,,.3                                                                                                                                                    --
eu.-;-        r,                                                                                                                                                                - c-
                                                                                                                                                                                                                                  ; p' '
                ---<-                                            s
* 1                                                                                                                      4
  , . ;. 4 ;
g                                                                                          g
                                                                                                                                                                                                                                                    '(        #
p'                      1
                                                ?M. .
1 y        1 i.i                                              't4                    kJ
        'sjf3                                        ,* . . ,-
                                                                                                                                                  )                                                                                                                                                                                                                                      4            [              7 y';(
                                                , ., a v.                          c,..h.f-        -
t
                                                                                                                                                                                                            ')I.'.-                    4 4
                                                ;%;_ .>: g? ;                                                                                                                                        u                                                                    'y                                                                            35g
                                                ',s n
s                                                                        <                      s s
s 7                  fh
            - T                    A                      ,,[,*                                      4 I                                                                                                                                                                      4
                                                                                                                                                                                                                                                                                                                                            'r,k.
s e, 4m .p.                                    m,                                3c                                                        ,
I                                              -
v .9,            _
r--
                                ~.N?'                    f.'f s
                                                                                                                                ,            ' ,l:
r-              ks                1,                                                                                                                                                                                                                                    ,
                                                                                      -f C 43                                                                K; ' '                                                                                                                        +
j'                                            + l If T
l N,"                                                          ; '                                                        I                                                                                                                                                                                                                                    '*CL                                        I''
r                          >
4,'f- . t_                                                      ,
g ."
1 t ~4 -          -
                                                            -                                -t r <-x.                '
                                                                                                                                                                                                                                                                                                                                                                                                            , ;G .r
_z_,                                                            y k ' I - 'l1                                        ,I                  w                p f
                                      ,- 75C      - ;,
i' ge                                                                                                                                                                [                                                    3
:        x' 4
a                                                                                                                        . , , a,
                        ^5 4      _I                                          fh;                                                                                                                      ',
[' % ,
t            *_*-('
q
_i ,
t t              ,
1'4*, , .
(v I
l 4 r
                                                                                                                                                                                                                                                                                                                                                                                                                    ''l e,L sge                                                    .
4
                                                                                                                                                          -                                                                                                                                                                                    +
            $                                  > f _' * ".'
                            ..,y. . . .                                                                                                                                        e                                                                                    !
1'                                                NO                                                                                                                                                                                                                                                                                                                            i                      , , -
                .i y                                                                ''
t b)+.
r--
            = [ h,                                                                                '
                                                                                                                                                                                                                                                                                                                                                                                              ; . G',
4 y...%
                                                                                      ,_7,                                                                                                                                                                                                                                                                                                                          4 pj, a                                              ,-
              )                                                              '!<d''s.                                    7 4                                                                                ,,
                                                                                                                                                                                                                                                                                                                                                                                        \              l                a I                                                                                                                      'k''                                                                                                                                                    -S } 5(I t Q-                                                                          4                                                                    4 1        h*,'.,-
i            + Lj                    2 q.g n.
1                                                  . D ,. 4                                                                                                                                                                                                                                                                                                                                      ' n,1
              ~+                                                                                                                                                                                                                                          '
                                                . p;..g
_,n                    'k                        %
{                    $
                              ,,.8,&
: m. '.
* y                3 g                                                                        *- y                                                                          I
            . fl.,
L ' p,/                                                                                                                                                                                                                                                                                                                                                                                            (
        ,1,            5                    y .L g-                                                      w
                                              =''
g t                      A*'
0 s                s
                .,-,'.=,-,                                                    1 k
l' J":'  A e<                                      u                                                                                                                                                                                                                                                                                                                          ,
[Y '                  -
1 y
g-i V ' . .{ 4, .                  .
e                j 4
                                                                  =..I
                                                                    ,b                  *
                            ' Qy '. y                                                                                                                        '
                . ; m'3 {2-
                            ,f.
ipQ.A , &a                                                                                r o
s" e > _                    ,e
                                                .F p                %                                                    %
S,L
:'t,.
                      '?
P.
s                            ,
b
* l c. '                                      v k "j t t... ,. s+ . " . -
* n- .;                                                                                          ,
cia.:f--.                  ,
c.= . 3 _ , . 4 .
E                                            ;,) h-I          s                    t          ^l            '...
jh,'
                                                                                        'A,Z..,','' %
                                                      -s                                            '{-          '                            f ng                                      y                                          4._
                    '                    )
:        f.. g                                          g-
                  -'![_%'          ' .5                                                                                                                                                                                        i                y                                                                          -_
 
CESSAR mn"e no.
2.1        PLANT FAMILIARIIATION This first major task in this analysis was plant familiarization.
The objective of this task is to collect the information necessary for identification of appropriate initiating events, determination of the success criteria for the systems required to prevent or mitigate the transients and accidents (the front line systems) and to identify the dependence between the front line systems and the support systems which are required for proper functioning of the front line systems.          This task was primarily an information gathering task.
The    information    collected    in    this  task  included  design information, operational information and information on plant responses to transients. The design information for the systems within the scope of the System 80+ Standard Plant design is contained in the appropriate sections of CESSAR-DC.            The System 80+ Standard Plant design includes the standard NSSS systems, the front-line safety systems and the BOP and support systems.
The transient analyses in Chapter 15 of CESSAR-F UI provided the basic information on plant responses to accidents and transients.
.This information was supplemented with discussions with safety analysts and licensed operators and with information contained in the report, "Depr to NRC Questions @"gpurization              .and Decay Heat Removal - Response'lO operator actions during plant transients System 80 plants        .
wereevalg}ydandestablishedbasedonC-E'sEmergencyProcedure Guideline      and discussions with licensed operators at C-E and at the Arizona Nuclear Power Project.          Surveillance requirements and operability definit g were        derived from C-E's Standard Technical Specifications          and, where more specific detai g s needed,  from the Palo      Verde Technical Specifications Maintenance  information,    where needed, was based on common industry practices.
                                    )                                  PRA The  Reif D 6 19 "kE'1hejgy t          ,  several  other  pub g edguide, studies                          t NUREG-1335g      a,nd NUREG-1150g IDCOR        IPE Procedures were also  reviewed as part of the plant familiarization task.        The objectives of these reviews was to provide a broad overview of areas to be addressed in this analysis and to identify potential problem areas.
2.2        ACCIDENT BEOUENCE DEFINITION The second major task in this analysis was the accident sequence definition. The objective of this task was to qualitatively identify those accident sequences which lead to severe core Amendment r B-5                  December 15, 1989 I
 
l                                                                                                                                              1 CESSAR M Ara 1
damage. This was accomplished using event tree analysis.                                                        Event tree analysis involves defining a set of initiating events and constructing a set of system event trees which relate plant system responses to each defined initiating event. Each system event tree represents a distinct set of system accident                                                                                      ;
sequences, each of which consists of an initiating event and a                                                                              ,
combination of various system successes and failures that lead to                                                                          i an identifiable plant state.              Procedures for develogng system                                                                  j event trees are described in the PRA Procedures Guide                                                        .                              j 1
For this analysis, the small event tree /large fault tree approach
  .was used.          In this approach, only the front line systems which respond to mitigate an accident or transient are addressed on the event tree.            The impact of the support systems is addressed I  within the f& ult tree models for the front line systems.                                                                                  !
The first step in defining the accident sequences was to select                                                                            l the initial ~ set of initiating events to be addressed in the                                                                                '
analysis.          (Note:    initial selection of initiating event is                                                                      i considered          to be partg3gf the    plant familiarization task in the                                                                .
PSA Procedures Guide              ). A Master Logic Diagram (MLD) was                                                                      l constructed to guide the selection and grouping of the initiating events.        An MLD is. essentially a top level fault tree in which the general conditions which could lead to the top level event are deductively determined.          For this analysis, the top event on the MLD was defined to be "offsite release". The bottom level                                                                              l elements on the MLD established the initial groupings of                                                                                    ;
initiating event types to be considered.                                                                                                    I The next step was to develop an initial list of event initiators.                                                                          )
First,        the lists of suggestegventtheinitiators from the PRA Procedures G                    ,
were extragd PSA Procedures Guide                                  ,
the IPE Methodology Manual g              and the lists of event initiators and        final initiat gg event Cliffs IREP Study g ,
groups weTI57xtracted          , the Oconee PRA from the g vert                  and the Arkansas IREP Studytff8ki n PRA        These lists of event initiators were then condensed into a single list.                                      This list was then reviewed to determine if any event initiators could be eliminated as not applicable because of plant design features, or if event initiators had to be added to the list because of-new design features.          Events- initiated by failures in the support systems were of particular interest because of the potent.a1 impact on the responding safety systems. The support system designs were evaluated to determine if there were potential failures of the j  support system or one of its trains that could lead to a reactor l
trip (i.e., initiate a transient) and at the same time, disable a portion of one or more of the responding safety systems.                                                      If so,                      i the failure of that support system was considered to be a special                                                                          j.
Amendment F B-6                                          December 15, 1989 l
l  , . , - .                                                          - - _ -    - , . - _ , . - - . - - .                  . . . - .
 
                                                                                                                                  )
CESSAR M Lm.                                                                                                              l l
l 1
initiator and was kept distinct from the other transient                                                                  !
initiators. The event initiators were then grouped into initial l      initiatingeventclassesbasedonthebottomeventsongeMLD, transient grouping                                      in Chapter 15 the initiating events analyzed in the other PRAs oyg$g{ g '{8'19'2bh*    ,
and the special initiators as described above.
An iterative process was then used to select the final set of                                                            ;
initiating events and to define the ' event sequences.                                                First, an          '
initial draft of an event tree was developed for each of the initial initiating event classes based on plant system responses to the specific type of initiator. These event trees were then                                                            ,
compared and where the system responses with respect to                                                                  ;
preventing core damage were the same or equivalent, the classes l      were combined. The event trees were then briefly evaluated for the individual initiators within the class.                                                    If the system responses to the' specific initiator were not covered by the class                                                        ,
event tree, the initiator was either transferred to an event                                                              ,
class for which the system' responses were appropriate, or a new event class was created and a new draft event tree was developed.
i~
This process was repeated until a set of initiating event classes                                                    -
t were defined that included all the initiators in the original                                                            i list and the event tree for each evdnt class covered the system responses for each initiator. The final event trees were then                                                        '
prepared and the description and success criteria were defined                                                      -
for each element on the event trees.                                              In general, the success            4
'      criteria for the event tres elements were based on the system                                                            -
performag CESSAR-F      .
used                for the Chapter 15 and Chapter 6 analyses in For elements not addressed in the Chapter 15 and Chapter 6 analyses, transient analyses                                      for success criteria equivalent plants warg' gag on          other and the judgement        of              the                        transient      analysts    with    confirmatory thermgpydraulic                              trgsient                analyses      using  codes      such as CENTS        or MAAP                                          as appropriate.
The set of initiating events and the event tree structure for each initiating event also evolved as the plant system designs evolved. As design changes were made, the set of initiating events was reviewed to determine if there were any changes (additions - or . deletions) and the event trees were reviewed to                                                        ,
determine if any changes were needed due to changes in system responses or responding systems. The event tree structure also changed to reflect the truncation of sequences with very low probabilities.                The PRA was used as a design evaluation tool.
Thus, the event trees were quantified at several times to evaluate    the              impact of major design evolutions.                                            These quantifications reflected the following major design change                                                              -
groups:
Amendment F B-7                    December 15, 1989
 
CESSAR !!!h                                                                                                                ;
A. Incorporation of the front line system design enhancements.
B. Incorporation of component cooling water system changes.                                                              j C. Incorporation of the electrical distribution system design                                                            i changes.
At-each quantification step, low frequency sequences were tagged for truncation.                          These sequences were deleted only after determining that they were not impacted by other design changes.
The event trees presented in Section 3 represent the final System                                                          -
80+ design.
All transients require rsactor trip                                      for reactivity              control.
Failure of reactor trip leads to an Anticipated Transient Without Scram (ATWS). Because of the special nature of this type of an event, ATWS was treated as a separate initiating event with a specific event tree.                        Thus, failure of reactivity control was not included in the transient event trees.
2.3        SYSTEM MODELING Each system event tree, as described in Section 2.2, represents a distinct set of system accident sequences, each of which consists of an initiating event and a                                  combination of various system successes and failures that lead to an identifiable plant state.                                                            !
Quantification                    of      the    system    accident      sequences requires knowledge of the failure probability or probability of occurrence for each element of the system accident sequence. The initiating                                                            '
event frequency and the probability of failure for a system accident sequence element involving the failure of a single component can be quantified directly from the appropriate raw data using methods described in Reference 4.                                            However, if the system accident sequence element represents a specific failure mode for a system or subsystem, a fault tree model of the system or subsystem must be constructed and quantified to obtain the desired failure probability.                              Construction of the fault tree requires a complete definition of the functional requirements for the system, given the initiating event to which it is responding, and the physical layout of the system. The system fault tree is a graphic model of the various parallel and serial combinations of component giures failure mode that would result in the postulated system The symbols used in constructing the fault tree models are presented and defined in Figure B2.3-1.
The evaluation of each fault tree yields both qualitative and quantitative information.                            The qualitative information consists of the "cutsets" of the model.                                  The cutsets are the various Amendment T B-8                            December 15, 1989
 
CESSAR tilsnemo.,                                                                  !
l I
combinations of component failures that result in the top event, i.e., the failure of the system.      The cutsets form the basis of the quantitative evaluation which yields the failure probability                  J for the system accident sequence element of concern.                              )
The quantitative evaluation of the fault trees yields several                      I numerical measures of a systems failure probability, two of which                    '
are typically employed in the event tree quantification, i.e.,                    1 the unavailability and unreliability. The unavailability is the                  j probability. that a system will not respond when demanded. This value is used when the system accident sequence element                          ,
represents a system function or action which is performed quickly, such as the reseating of a previously opened safety valve, or if the element represents a particular condition, such                  '
as offsite power unavailable at turbine trip. The unreliability 1            is the probability that a system will fail (at least once) during a given required operating period.      This value is typically used when the system accident sequence element specifies a required operating period for a system, such as auxiliary feedwater system          ..
fails to deliver feedwater for four hours.      The unreliability is        7 usually. added to the unavailability when the system accident                    .
l            sequence element represents the failure of a standby system to actuate and then run for a specified period of time,                            t Two types of human failures are typically included in fault tree analyses. They are " pre-existing maintenance errors" and failures of the operator to respond to various              demands.        ,
Pre-existing maintenance errors are undetected errors committed              ~
since the.last periodic test of a standby system.      An example of          ,
this type of error is the failure to reopen a mini-flow valve                f.,
which was closed for maintenance.      A failure of the operator to respond includes the failure of the operator to perform a required function at all or to perform it correctly. An example of this type of error is the failure of the operator to back-up the automatic actuation of a safety system.
For this PRA, failure of the operator to respond to various demands where there was a time const            was quantified using
!            the Human Cognitive ReliabilityModelg            The human cognitive reliability model is a set of time dependent functions which describe the probability of a crew response in performing a task.
l The human cognitive reliability model permits the analyst to l
predict the cognitive reliability associated with a non-response for a given task or series of related tasks, once the dominant type    of  cognitive  processing (skill-based, rule-based or knowledge-based), the median response time for the task or tasks under nominal conditions and performance shaping factors such as stress levels or environment are identified. The inherent time dependence in this model made it ideal for evaluating operator Amendment F B-9                December 15, 1989 l
  ,  , , ~ _ -              ,_ _  y                                                  -
 
CESSAR !!nhwa responses    during  a  transient. The  failure  probability    for  i o  " pre-existing    maintenance    errors"  was g antified      using  the Handbook of Human Reliability Analysis                  The Handbook    of
~
Human    Reliability    analysis    is  an  extension    of  the  human  i reliability analysis Ighodology Reactor safety Study developed for WASH and is intended to provide 1400, the methods, models and estimated human error probabilities to enable analysts to make quantitative or qualitative assessments of the occurrence of human errors that affect the availability or operational reliability of engineered safety systems and components.              The emphasis is on tasks addressed in the Reactor Safety Studyt calibration, maintenance and selected control room tasks related to engineered safety features availability.            It is the best available source for evaluating human performance with respect to maintenance,    calibration,  testing and other tasks performed i
during normal plant operation.      However, its time dependent model is not as thorough and explicit as that provided by the human cognitive reliability model.
For this PRA, the small event tree /large fault tree approach was selected.      The event trees developed for this PRA address the i  response of the front line systems, that is, those systems                  >
1  directly involved in mitigating the various initiating events.
The support systems, basically the electrical distribution and component cooling water systems, were fully modelled within each of the front line system models. As discussed in Section 2.2, failure of the support systems have the potential for initiating            ,
a transient and, simultaneously, disabling a portion of one or more front-line safety systems.            Separate front-line safety system fault tree models were prepared for evaluating the event trees for these special initiators.          These special front-line safety system fault tree models reflected the specific impact of the initiator on the safety system and other support systems.
For example, loss of one division of component cooling water would result in an equipment protection reactor trip. It would also result in loss of cooling for one diesel generator, one motor-driven emergency feedwater pump, two high pressure injection pumps, one containment spray pump and one shutdown                -
cooling pump. Therefore, for the initiator, " Loss of one CCW/SW Train", the front-line safety system models were modified to show that the above equipment was not available to respond to the transient.
Amendment F B-10                  December 15, 1989
 
                                                              ~
I 1
1 Or Gate            Output event occurs if one or                    l more of the input events occurs b                                                                      :
m And Gate          Output event occurs if all of the                .
input events occur.                              !
l      r3                                                                      -
l e    i Basic Event        Basic Fault event requiring no              -
further development.                            ,
I'        Developed Event  An event which is described by a              j fault tree model developed inde-pendently. Typically a set of com-o                          ponents in series which can L                          always be treated as a unit.
L                                                                            ,
[
Transfer In      Used as a method of convenient-i ly segmenting the tree for draft-                -
i                          ing purposes and to avoid L                          duplication of portions of the l
tree indicates continuation to                    ,
other portions of the tree.
L                                                    Amendment F December 15, 1989 i
Figure I
jg7            Fault Tree Symbology                        B2.3 1
 
                                                                                                                                                                ..                            ~ '
N~
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                                                                                                                                                '                                                            ,  4
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I CESSAR tilW,cu..                                                                                            l 2.4          DATA ASSESSMENT i
Reliability data is needed for the quantification of the system                                              l fault trees and the system accident sequences which result in                                                l severe core damage.        The data needed for this quantification                                          )
l includes:
A. Initiating event frequencies.
1 B. Component failure rates (demand and time-dependent).
C. Component repair times and maintenance frequencies.
D. Common cause failure rates.
E. Human failure probabilities.
F. Special event probabilities            (e.g.,        restoration of offsite power) .
G. Error factors for the items above.
Generic reliability data was                used g        this analysis per the guidance in the PSA Procedures Guide                    .      The primary source of data    used    in  this  PRA    was      the  "PRA            Key Assumptions and                      :
                                                                                                              ~
Groundrules" (KAG) documenk5f Appendix                  A to Chapter 1 of the EPRI ALWR Requirements Document              ). Other industry-accepted generic data sources were used, as needed, to supplement the data in the KAG.
The basic initiatiny3pventEPRI  frequenclygg)were NP-2230                  and  extracted the NREPfrom        the Generic PSA    Proceg Data Base Guide      ,
initiating frequencies, presented in the Oconee PRA ggpt and the Calvert Cliffs IREP l  the ~ ZigRA(15)      , The were also used as guidelines. The appropriate basic Report      ,
initiating event frequencies were used to calculate the needed initiating event class frequencies.
The basic component        failure rate data and associated error factors was extracted from the KAG, which contains a compilation of generic failure rate data from other nuci                                                This datag supplemented witgta from                  WASH 1400g) sources.the and the NPRDS data base as g Data Base      ,    IEEE Std. 500          ,
needed. Component maintenance frequencies and repair times were l
calcu      d using the procedures outlined in the PSA Procedure Guide i
Common cause failures of components were explicitly modeled in the    system fault trees.              The common              cause        failures      were The appropriate common
!  calculated using equivalent Beta factors.
l l
l Amendment r l                                          B-11                                December 15, 1989
 
CESSAR HWic.ra E
cause failure rate f actors were taken dire::tly from the KAG or              !
were    quantified      using  a    process      equivalent    to  that i    data extracted from other data            i outlinep33g4,ge3 6,@,3 D ,thas appropriate.
sources
(
As  discussed    in    Section  2.3,      two types of human failure; pre-existing maintenance errors and failure of the operator to                ,'
perform varhous actions during a transient; were modeled in this analysis. Pre-existing maintenance errors were evaluated using            ,
the met g y described        in the Handbook of Human Reliability Analysis      ,  and the operator responses during an event g e modeled using the Human Cognitive Reliability Model                      .
l    Quantification    data was prgily extracted from the Handbook of Human Reliability Analysis          .
l l
The methods and data used for quantifying the probability of special events (e.g., the restoration of offsite power within a              I given time period) were dependent on the specific event.                      .
1  3.5        ACCIDENT SEQUENCE QUANTIFICATION                                  .
The basic objective of accident sequance quantification in to generate the core damage cutsats and calculate the core damage frequency due to internal events for the System 80+ plant. The n    total. core damage frequency, due to internal, events, is the sum frequencies of the system level accident sequence of the frequencies for those accident sequences which result in core damage. As described in Section 2.2, the system level accident sequences leading to core damage were identified using event tree analysis. Each system level accident sequence consists of an initiating event and one or more additional elements, each representing either a front line system failure or a special event such as failure to restore offsite power within a given I    time or the most reactive rod sticking out of the core following a reactor trip.
I The core damage frequencies for the system level accident sequences were determined using a six step process. The first step was to calculate the frequencies for the initiating events and the special events and to represent these elements as single              ;
event fault trees (an OR gate with a single input) . The second step in the process was to evaluate the front-line safety system              ;
fault tree models to generate the system cutsets and calculate the system unavailability both for the base case and for all applicable special cases. (See section 2.4 for a discussion of
    -spec g cases.)      The beta draft version of the IRRAS 2.0 computer code      was used to build and evaluate all fault tree models needed for this analysis.
Amendment F B-12                    December 15, 1989
 
CESSAR tilbn..                                                                                        ;
i The next step was to generate the initial core damage cutsets for                                      ;
each sequence leading to core damage.              This was done using . full                          ,
l fault tree linking to combine the system cutsets for each element in the sequence.      This process also yielded an initial point estimate of the core damage frequency attributable to each core                                        i damage sequence. The beta draft version of IRRAS 2.0 was also                                          -
used for the sequence evaluations.
The fourth step in the accident sequence quantification was the                                        '
recovery analysis.      For core damage sequences with a core damage                                  ,
frequency point estimate        of less than 1.0E-12, the recovery analysis consisted of reviewing the core damage sequence cutsets                                      j to identify and remove all nonsense cutsets. (e.g. a cutset containing a demand failure and a run failure for the same
, component is a nonsense cutset.)        A full recovery analysis was
! performed for each core damage sequence with a core damage l frequency contribution point estimate of greater than 1.0E-12.
First, the sequence cutsets were reviewed to identify and remove                                      ;
all nonsense cutsets.      Next, each remaining core damage cutset                              ..,.
in the sequence was evaluated to determine if there were any
* j appropriate recovery actions.      The types of recovery actions that                                ;
war,e considered included:
A. Recovery of offsite power.
B. Aligning the standby    AC source to a vital 4.16KV bus.                                    ,
l l C. Operator actions to realign a system in order to bypass a                                        ,
failed component.                                                                            }  !
D. Operator actions to align an alternate system to perform a safety function.
E. Operator actions to start a system that failed to start automatically as required.
F. Operator actions to open manual isolation in a discharge                                        ;
path or to close manual diversion (bypass) valves.
G. Repair of failed components where deemed appropriate.
The evaluation of the recovery actions included an assessment of the time needed to perform the recovery action and the time available to perform the recovery action.                      When one or more recovery . actions were - deemed appropriate and effective for the                                    '
fault represented by the cutset and that there was sufficient time to identify the need for and to perform the recovery action (s),  a  recovery action element for each appropriate recovery action was added to the cutset.                Each recovery action element represented failure to perform the indicated recovery action.
Amendment F B-13                              December 15, 1989
 
i l
CESSARtmLm.
l l
l l
When the recovery analysis was completed for a given sequence, the sequence was requantified and an uncertainty analysis was performed. When the recovery analyses for all sequences for a l      given initiating event were completed, all of the cutsats for all i      of the sequences were combined into a single cutset file with all L      duplicate and non-minimal cutsets removed.                                      This " initiating o
event      sequence"    was    quantified          and an        uncertainty              analysis l      performed.        The result          was    the  mean        core        damage        frequency l      contribution        for the            initiating event              and the associated uncertainty measures.
When the recovery analysis and requantification was completed for l      all sequences, all core damage cutsets for all sequences for all initiating events were combined in a single cutset file which was then requantified as a single model and an uncertainty analysis was performed.          This unified model was used to identify the i
dominant core damage cutsets regardless of initiating event and to evaluate component importance measures taking all initiating events into account.
2.s            EETERNAL EVENTS ANALYSIS External events are, typically defined as those events that result in a plant perturbation / transient, but are not initiated by the                                                  .
plant systems.        The definition of external events has evolved over time. Many external events, such as earthqua'kes and severe storms are truly external to the plant. Other events typically included in the external event class, such as a fire in electrical cabling - or flooding from a ruptured pipe, actually occur inside the plant.                Some of these external events not only initiate a plant transient, they also introduce complications which may negate mitigation features and/or cause additional common cause failures.
Past      PRAs  which have          addressed      a:rternal        events (15,16,20,40) identified potential plant vulnerabilities to these events which were found to make significant contributions to overall public risk.      Therefore, external events were addressed as part of the                                              i System 80+ PRA.          The external events evaluation was performed in two stages.        The first stage was a qualitative evaluation of the external events to determine which events would require detailed quantitative evaluation.                The second stage was the quantitative evaluation for those events identified in the first stage.
The external events qualitative evaluation involved the following four steps:
A.        Exte{gSf16'N'N)j            were      identified        by              reviewing      past PRAs                  g ndand  PRA    guidance documents such as the g Procedures guide                    the PRAT Fundamentals document prepared by BNL.
Amendment F B-14                                    December 15, 1989
 
i l
1 C E S S A R t m i r.c a r..
l l
B.              Events with similar plant effects and consequences were grouped together.                                                                                                ,
C.              Criteria were established to determine which external events                                                    (
were    insignificant risk contributorE and thus could be excluded from detailed quantitative evaluation.                                                    The        ;
screening      criteria were based on design requirements yt forth in the EPRI ALWR Utility Requirements document                                                ,
generally accepted regulatory pr g cesconsiderations,                        as documentedsiting in the          ,
NRC Standard Review Plan (SRP) considerations and frequency of occurrence.
D.              Each external event identified in step 2 above was then evaluated against the screening criteria established in step 3 to determine whether detailed quantitative analysis was needed. This evaluation also considered the insights gained                                                    '
5,#Y,N,40gf PRAs for present generation power 1
ts                  ,
Batsed on the qualitative screening, only two external events,_
tornado strikes and earthquakes, were selected for detailed quantitative analysis. The primary impa'ct of a tornado strike is                                                          ,
an extended, non-recoverable loss of offsite power.                                                        Therefore,      2 the tornado strike analysis used the standard event tree / fault tree analysis methodology described in Sections 2.2, 2.3, and i                                                                                                                                      '
2.5.                Evaluation of the impact of earthquakes on public risk requires special methodology.                                The seismic evaluation methodology and results are described in detail in Section 4.                                                                            <
4 2.7                    CONTAINMENT RESPONSE ANALYSIS (LATER) 2.8                    CONSEQUENCE ANALYSIS (LATER) l 1
L Amendment F B-15                                  December 15, 1989
 
CESSAR inlinc no.                                                                                                    .
2.9                &M&LJ8IS GROUNDRULEE                                                                              ,
Appendix A. to Chapter 1 of the gI Advanced                      containsLight                a Water  Reactor key Utility            Requirements Document                                                          set  of assumptions and groundrules for performing a Probabilistic Risk Assessment for an advanced light water reactor.                                                The groundrules used for the System 40+ PRA are:
A.          Only events with potential for core damage were addressed.
B.          Only initiating events that occur at nominal full power were
* considered.                                                                                              t C.          Where needed, realistic, best-estimate assumptions were used                                            !
when evaluating plant responses to an initiating event.                                                  ;
D.          Core damage was assumed to have occurred if the collapsed fuel level  in the  reactor in the core was uncovered and had decreased a  such temperature    that        the of active,F 2200    or  '
higher was reached in any node of the co}}

Latest revision as of 04:38, 21 December 2024

Amend F to C-E Std SAR - Design Certification (CESSAR-DC)
ML20011D517
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Issue date: 12/15/1989
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