ML20005C038
| ML20005C038 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/13/1981 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-660, RTR-NUREG-737, TASK-1.A.2.2, TASK-1.C.5, TASK-2.D.3, TASK-TM NUDOCS 8111180307 | |
| Download: ML20005C038 (101) | |
Text
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x Commonwealth Edison
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) one First Nat:onal Plaza. ^hicago. Illinois
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Novembe r 13, 1981 ihh Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Responses to FSAR Questians NRC Docket Nos. 50-454/455 and 50-456/457
Dear Mr. Denton:
This is to provice aavance copies of answers to questions from the NRC Staff regarding the Byron /Braidwood FSAR Attachment A to this letter contains responses to Questions 130.06 and 110.66.
Also contained are three items from the FSAR Appendix E, TMI Action Plan.
These are a revised response to NUREG 0737 Item I.C.5 and responses to NUREG 0737 Item II.D.3 and NUREG 0660 Item I.A.2.2.
All of this naterial will be incorporated into the Byron /Braidwood FSAR in future amendme.its.
Fif teen (15) copies are provided now for your early review and approval.
One (1) signed original and fifty-nine (59) copies of this letter are provided.
Please address any questions regarding this matter to this office.
Very truly yours, jf QLw n T.R.
iramm Nuclea r Licensing Administrator Attachment
$0 0 s
i 2876N
.g p,
8111180307 81111'3 PDR ADOCK 05000454 A
1 ATTACIIMENT A A.
List of Byron /Braidwood Questions Addressed i
110.66 130.06 B.
List of TMI-Action Plan Requirements Addressed g pendix E Items)
I.C.5 (Revised)
II.D.3 I.A.2.2 from NUREG-0660 e
e B/B-FSAR QUESTION 110.66
- Provide an assessment of those piping systems, components, and supports identified above for the new seismic response spectra.
If the applicant chooses to use the Marble Hill spectra for its evaluation of piping and supports, its use must be justi-fled with respect to the structural differences at critical sections of the Byron /Braidwood and Marble Hill plants.
a.
For piping, provide a table showing for selected critical locations, the design basis seismic stress, the new seismic stress, the total stress (pressure, weight, and seismic),
and the allowable stress, b.
For valves, provide a similar table showing the acceleration values for the design basis, the new seismic design, and the allowable accelerations.
- i c.
For supports, provide a similar table showing the support loads for the design basis, the new seismic design, and the support allovrble load.
d.
Provide a qualitative measure of the overall margin to failure for each of the above components.*
RESPONSE
As the Marble Hill design basis is replicate to the Byron and i
Braidwood design and had implemented Regulatory Guide 1.60 spectra, it was used in the assessment of the Byron and Braidwood piping systems, components and supports.
The structural differ-ences at critical sections of the Byron /Braidwood and Marble '
Hill plants are as follows:
i Methodology I
Based on the Marble Hill design, the stress level in structural elements unique to Marble Hill as compared to Byron /Braidwood were identified.
The overstressed elements were reviewed against the actual average material strength. Reassessed elements having a Safe Shutdown Earthquake (SSE) stress leveli less than the yield limit of the material will be considered acceptable due to the localized nature of the forces.
Elements still overstressed will be reviewed on a case by case basis in order to determine their impact on the safe shutdown of the plant.
i l
l Q110.66-1
B/B-FSAR Reassessment of Safety Related Structures Containment Building Structure The initial assessment of the containment structure was based on a comparison of the Byron /Braidwood design with that of Marble Hill.
Areas where Marble Hill had unique design features were identified and force comparisons
.. - - tabulated.
These tabulations are shown on Table 110.66-3.1.
The Marble Hill and Byron /Braidwood design bases, however, were based on conservative assumptions as is appropriate for initial design purposes and, therefore, the force comparison does not reflect the adequacy of the structure.
Subsequent to the initial assessment, the following conservatisms were identified and removed and the structure reanalyzed.
a.
The overall containment overturning moment, axial force and shear used in the reassessment were obtained using a shell model rather than the beam model previously used.
b.
The stiffness of the elements used to represent the reactor cavity wall was modified to account for initial concrete cracking due to shrinkage, thermal effects and restraint.
c.
Full hydrostatic uplift forces included in the designi basis basemat analysis were excluded from the present analysis.
This is a more realistic assumption due to' the uplift of the basemat during a seismic event.
d.
The factor applied to the effect of a single horizontal excitation for basemat analysis to account for the effect of three components cf excitation was reduced to 1.05.from the 1.10 used'in the' design basis analysis.
The factor 1.10 is based on a study performed during the design bases analysis and it included 5 to 8% safety margin.
The results of the containment structure assessment based:
on the refined analysis show no overstress in the contain-ment structure.
Containment Internal Structures Concrete No unique design features were located.
0110.66-2
o i
B/B-FSAR Steel Columns No unique design featires were located.
Steel Beams Eighty-three beams of approximately seven hundred and forty per unit have a unique design due to seismic forces.
These beams were assessed against the Marble Hill SSE response spectra and the results tabulated as follows.
Acceptance Criteria Number of Beams Lesser of 0.95y or 1.6 AISC Allowable where 75 Fy=36 KSI Lesser of 0.95 Fy or 1.6 AISC Allowable where 6
Fy=42.3*KSI 1.0 Fy where Fy=42.3*KSI 2-1F3
- Actual average material strength for A36 structural steel.
Auxiliary / Fuel Handling Building Basemat The results of the assessment are tabulated as follows.
Acceptance Criteria Number of Mat Areas (Finite Elements)
ACI Allowable F'c=3,500;Fy=60,000 10 ACI Allowable F'c=5,265*;Fy=67,000*
_7 17
- Actual average material strength for concrete and Rebar.
Q110.66-3
B/B-FSAR 4
Shear Walls The results of the assessment are as follows.
Acceptance Criteria Number of Walls Vertical Horizontal Steel Steel ACI (0.9Fy)
F'c=3,500; Fy=60,000 45 64 ACI (o.9Fy)
F'c=3,500; Fy=67,000*
19 0
j 1.0 Fy F'c=3,500; Fy=67,000*
1 0
- 1. 03 5 Fy (M.F.=0. 9 66)
F'c=3,500; Fy=67,000*
O 1
65 65
- Actual average material strength for Rebar.
Concrete Beams and Slabs No unique design features were located.
Steel Columns The results of the assessment are as follows.
3 Acceptance Criteria Number of Columns AI.9C Allowable Fy=50 62 AISC Allowable Fy=56*
4 l
68
- Actual average material strength for A577 structural steel.
Q110.66-4
B/B-FSAR Steel Beams The results of the assessment are as follows.
Acceptance Criteria Number of Beams Lesser of 0.95 Fy or 1.6 AISC Allowable 189 Fy=36 Lesser of 0.95 Fy or 1.6 AISC Allowable 52 Fy=42.3*
1.0 Fy Fy=42.3*
1 242
- Actual average material strength for A36 structural steel.
Q110.66-S
B/B-FSAR RESPONSE - 110.66a In reassessing the piping systems, all the piping subsystems or problems were identified and grouped according to system as shown in Table 110.66-1.
From the 319 piping problems listed, a representative sample of 40 piping problems were selected using the following criteria.
1.
Selected all piping problems with level C design basis levels " Equation 9 NB/NC - 3600" greater than 80%
of the allowable limits.
2.
Selected piping problems ranging in size from 3/4" to 48" nominal diameter.
3.
Selected piping problems from a range of building elevations to provide a diversity of response spectra input.
4.
Selected at least one piping problem from each system required for cafe shutdown.
Methodology In assessing the piping problems, the problems were analyzed for at least 30 modes or a range of frequency covering 33 Hertz.
The damping valves used were in accordance with Reg. Guide 1.61 and the resultant stresses for service level C were compared against the allowable stress limits of Articles NB/NC - 3600 of ASME PBC Code Section III.
Table 110.66-2 contains a com-parison of the highest stress points in a given piping problem, the function of the line, location, elevation and allowable stress limits.
0110.66a-1
6 TABLE 110.66-1 BREAKDOWN OF PIPING SUBSYSTEMS System Number of Piping Problems Main Steam System 5
Main Feedwater System 18 Auxiliary Feedwater System 15 Emergency Diesel Generator 20 Component Cooling System 64 Essential Service Water System 52 Chemical and Volume Control System 70 Borated Water System 3
Residual Heat Removal System 20 Reactor Coolant System 30 Chilled Water System 22 TOTAL 319
~
Q110.66a-2
3 TABLE 110.66-2 l
STRESS LEVELS FOR ASME CLASS PIPING Level C Eq. 9 Stress (1)(KS1)
Code Oldg.
Pipe Design Ma rb le Level C Stress Subsystem Line Class Elevation Element Basis Hill Limit (2) (KS1) 1AF-02 Aux. Feedwater 2
Aux /364 Reducer 20.5 19.5 27.0 Pump Discharge 1AF-05 Aux. Feedwater 2
Aux /383 Elbow 16.4 16.8 27.0 Supply to Steam Generator 1AF-06 Aux. Feedwater 2
Aux /364 Valve End 11.9 10.3 27.0 Pump Discharge IAF-07 Aux. Feedwater 2
Aux /401 Valve End 21.3 22.5 27.0 Supply to Steam Generator E
1CC-42 Component Cooling 2
Aux /383 Elbow 7.5 5.4 27.0 4
fle t u rn Piping U'>N 1D0-16 T rans fe r Pump 2
Aux /383 Str. Pipe 5.9 6.0 27.0 Discharge 100-17 Trans fer Pump 2
Aux /383 Eq. Conn.
14.3 15.4 27.0 Supply IFW-12 Main Feedwater 2
Aux /383 Struct.
19.6 19.7 27.0 Valve Bypass Line Anchor 1FW-16 Main Feedwater 2
Aux /383 Valve End 14.8 12.7 27.0 Valve Gypass 1SX-32 Essential Service 2
Aux /346 Elbow 10.6 10.5 27.0 Water Return Headers (1)
Operating Pressure & Deadweight & SSE (2) 2.25 S fu r Class 1,
1.8 S f r n n-class 1, Defined at Operating Temperature m
h Q110.66a-2a
r TABLE 110.66-2-continued
' STRESS LEVELS FuH ASME CLASS PIPING Level C Eq. 9 Stress (1)(KS1)
Code Oldg.
Pipe Design Ma; b le Level C Stress Subsystem Line Class Elevation Element Basin Hill Limit (2) (KS1) 1 S X- 05 Essential Service 2
Aux /383 Elbow 14.3 15.2 27.0 Water
\\,
1SX-13 Essential Service 2
Aux /401 Elbow 11.8 12.5 27.0 Water Return Piping ISX-34 Essential Service 2-Aux /401 Struct.
26.1 25.7 27.0 Water Return Anchor Piping ISX-63 Essential Service 2
Aux /401 Struct.
19.2 20.5 27.0 Water Supply Anchor Ep Piping g
a 2SX-03 Essential Service 2
Aux /401 S t, r u c t.,
6.3 6.6 27.0
]
Water Return Anchor Piping I'
IWO-32 Chilled Water 2
Aux /451 Eq. Conn.
7.3 9.9 27.0 (1)
Operating Pressure & Deadweight & SSE (2) 2.25 S, r'o r C l a s s 1, 1.8 S f r e.ar-class 1, De fined at Operating Temperature h
0110.66a-3
TABLE 110.66-2-continued STRESS LEVELS FOR ASME CLASS PIPING Level C Eg. 9 Stress (l)(KSl)
Code Oldg. &
Pipe Design Ma rble Level p Stress Subsystem Line Class Elevation El ement Basis Hill Limit (2) (KS1)
Reactor Coolant Loop 1
Cont.
Hot Leg Elbow 31.0 29.7 39.6 Crossover Leg Elbow 32.3 29.8 39.6 Cold Leg Wel d 36.1 26.3 39.6 IMS06 Main Steam 2
Cont.
Elbow 15.7 15.5 27.0 386'-
464' 1 FWO4 Feedwater 2
Cont.
Elbow 8.4 8.8 27.0 390'-
407' RHR/SI l
Cont.
Elbow
- 45. 0 (3b5.6 38.0 3El 1RH02 System 393' Tee 39.1 27.8 38.0 G3 2
Anchor 17.1 31.0 33.3 43 1RC21 Sample 2
Cont.
Socket ta 393' Weld 13.4 10.7 28.5 Elbow 13.4 8.2 28.6 1CV02 Cha rging 1
Cont.
Elbow 24.1 25.4 38.3 393'-
413'
~
2 Wel d 24.2 19.7 33.3 1CV06 CVCS 1
Cont.
Elbow 22.0 21.7 38.3 393'-
414' 1CV09 Drain 1
Cont.
Weld 38.0 35.6 38.0 3/9' 2
Weld 20.8 25.2 28.6 (1) Operatin9 Pressure & Deadweight & SSE (2) 2.25 S for Class 1,1.8 S f r non-class 1, Oefined at Operating Temperature m
h (3) Designed to Level D stress Limit (=51 KS1)
Q110.66a-4
.. ~.
TABLE 110.66-2-continued STRESS LEVELS FOR ASME CLASS PIPING Level C Eg. 9 Stress (l)(KS1)
Code Bldg. &
Pipe Design Marble Level G Stress s
Subsystem Line Class Elevation Element Basis Hill Limit (2) (KS1) 1RYO5 Pressurizer 1
Cont.
Weld 15.4 27.4 37.6 i
~
Surge 393' Elbow 18.5 24.2 37.6 1CV03 Charging 1
Cont.
Elbow 18.8 17.2 38.3 2
393'-
Weld 8.0 12.3 28.6 413' ICV 36 RCP Seal 1
Cont.
Straight 51.9(3)41.6 45.0 Water Inj.
396' Pipe 2
Weld 23.3 21.2 33.7 di ICC21 Component 2
Cont.
Branch 25.0 20.4 27.0 G) 1CC22 Cooling 416' Weld 26.3 21.3 27.0 ICC23
]
M 1SX06 Service 2
Cont.
Water 403' Tee 9.8 22.1 27.0 Straight 5.5 23.1 27.0 Pipe Elbow 10.1 18.7 27.0 1SX08 Service 2
Cont.
Branch 34.0(4)34.7 2 7. 0 --
Water 403' Weld 25.1 25.5 27.0 Elbow 21.8 21.6 27.0 15003 SG Blowdown 2
Cont.
Weld 26.3 21.0 27.0 388'-
Straight 22.3 16.0 27.0 408' Pipe t
i (1) Operating Pressure & Deadweight & SSE (2) 2.25 S for Class 1,1.8 S for non-class _1, Defined at Operating Temperature m
h (3) Designed to Level 0 stress Limit (= 60'KS1) 4 (4) Designed to Level D stress Limit (= 36 KS1) t 0110.66a-5
l t
TABLE 110.66-2-continued STRESS LEVELS FOR ASME CLASS PIPING j
Level C Eg. 9 Stress (1)(KSl)
Code Bldg. &
Pipe Design Marble Level Q Stress S_ubsystem Line Class Elevatior Element Basis Hill Limit (2) (KS1) u
~
1CV60 Ercess 1
Cont.
Weld 26.4 22.4 38.3 Letdown 415' 2
Weld 18.7 22.6 28.6 i
ICC25 Component 2
Cont.
Weld 14.7 20.6 33.3 ICC28 Cooling 396' Weld 23.5 19.3 33.3 i
1S 004 SG Blowdown 2
Co nt.
Straight 18.4 14.1 27.0 i
387'-
Pipe i
401' 15I02 Safety Injection 2 Cont.
Straight 9. '.
10.9 28.6 i
.420' Pipe 15I04 Accumul ator 1
Cont.
Tee 30.5 29.4 45.0 k
System 393'-
p t
1 429' j
2 Elbow 21. 2 17.6 33.3 7
j 1SI09 Accumulator 1
Cont.
Tee 29.4 31.1 45.0 System 393'-
7b i
429' 2
Weld 22.5 17.6-33.3 r
1 l
I (1) Operatiiig l'ressure & Deadweight & SSE
{
(2) 2.25 S f r Cl as s 1,. l. 8 S f r n n-class 1, Defined at Operating Temperature m
h i
Q110.66a-6
. = _ -
/
8/B-FSAR RESPONSE - 110.66b A total of 89 valves were located in the 40 selected piping subsystems.
These valves were assessed as follows.
i 1.
61 valves had to meet allowable acceleration limits in each mutitally perpendicular direction (X,Y,Zi.
P 28 valves had.to meet. allowable--acceleration limits 2.
i in horizontal and vertical directicn.
Table 110.66-3 contains a comparison of the in line valve accelerations using the design basis and Marble Hill response spectra for the SSE event.
The valve types, sizes and allowable accelerations.are also shown on Table 110.66-3.
J l
i 4
1, i
1 4
i f
4 4
i Q110.66b-1
1 TABLE 110.66-3 l
VALVE ACCELERATIONS ON ASME PIPING VALVE DESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SUBSYSTEM SIZE / TYPE H
H V
H H
v H
H V
y 2
y 2
i 2
j lAF-0 2 3" Control ( A.0)
.564 1.333
.955
.562 1.185
.718 3.0 2.5 3.0 3" Control (A.0)
.906 1 264 1.059
.679 1.090
.774 3.0 2.5 3.0 i
1 3" Control (A.O) 1.393 1.575
.992 1.061 1.272
.753 3.0 2.5 3.0
.833f
.8C9 l
3" Control (A.0) 1.545 1.057 1.211 1.029 3.0 2.5 3.0 1AF-05 6" Check 1.718 1.836 1.001 1.661 1.857 1.042 3.0 2.5 3.0 6" Gate 1.406 1.857
.756 1.436 1.921 0.744 3.0 2.5 3.0 6" Control 1.459 1.983
.566 1.520 2.051
.504 3.0 2.5 3.0 6 " Ga te 1.578 2.111
.777 1.633 2.177,
.675 3.0 2.5 3.0 6" Check 1.354 1.073 1.853 1.307 1.194 l.955 3.0 2.5 3.0 4" Globe
- 1. 19 8
.987 1.263 1.195 1.093, 1.271 3.0 2.5 3.0 h-4" Check 1.315 1.534 1.269 1.250 1.561 1.302 3.0 2.5 3.0 4" Globe 1.678 1.467
.887 1.592 1.370 896 3.0 2.5 3.0
,1 1AF-06 6" Check
.559
.573 1.183
.506
.573 1.212 3.0 2.5 3.0 6" Gate (M)
.569 1.149 1.149
.516
.580, 1.160 3.0 2.5
_3.0 4" Control (A.0)
.500
.634
.922
.455
.626
.777 3.0 2.5 3.0 6" Gate (M) 1.216 1.005 1.561 1.098
.924 1.269 3.0 2.5 3.0,
I 4" Nozl Check
.956
.869 1.539
.897
.829 1.365 3.0 2.5 3.0 4" Globe (MO) 1.143
.997
.882
.883
.973
.780 3.0 2.5 3.0 4" Nozl Check 1.102
.452
.914
.931
.436
.671 3.0 2.5 3.0 4" Globe Iso (MO) 1. 3 41
.448
.876 1.034
.452
.635 3.0 2.5 3.0 1
1AF-07 6" Check 1.109
.953
.869
.960
.846
.670 3.0 2.5 3.0 6" Gate (M) 0.833
.717
.911
.726
.649
.728 3.0 2.5 3.0 6" Control (AO)
.598
.481
.851
.530
.430
.646 3.0 2.5 3.0 6" Gate (M)
.714
.570 1.117
.624
.501
.825 3.0 2.5 3.0 Q110.66b-2
TABLE 110.56-3-continued VALVE ACCELERATIONS ON ASME PIPING VALVE DESIGN BASIS SSE MARBLE HIuL SSE ALLOWABLE SSE SUBSYSTEM SlZE/ TYPE H
U V
H H
Y U
U Y
y 2
y 2
1 2
1AF-07 6" Check 1.126
.880 2.444
.924
.710 1.534 3.0 2.5 3.~ 0 4" Globe (MO) 1.702 1.103 1.338 1.035
.787
.865 3.0 2.5 3.0 4" Check
.856
.641
.936
.719
.547
.685 3.0 2.5 3.0 4" Globe (MO) 1.568
.663
.855 1.171
.564
.622 3.0 2.5.
3.0 1CC-42 3" Gate (M)
.760
.375 1.034
.535
.339
.608 2.12 2.12 2.0 3" Gate (M)
.430 1.943 1.470
.816 1.051
.798 2.12 2.12 2.0 1DO-16 3" Relief
.478
.853
.657
.497
.841
.600 lDO-17 3" Gate
.277
.347
.825
.293
.341
.608 3.0 2.5 3.0
~
3" Gate
.258
.312
.816
.289
.324
.592 3.0 2.5 3.0 3EL 3" Gate
.945
.453
.857
.819
.412
.739 3.0 2.5 3.0 f
91 3" Gate
.289
.477
.812
.315
.4 19
.623 3.0 2.5 3.0 g
1FW-12 3" Check 1.499 1.220 1.377 1.303 1.071 1.151 3.0 2.5 3.0 3" Control
.987
.857
.990
.874
.789.
.803 3.0 2.5 3.0 3" Gate 1.482 1.229
.696 1.232 1.066
.5 17 3.0 2.5 3.0 3" Ga te
.886
.849
.561
.763
.744 1.465 3.0 2.5 3.0 3" Control
.476
.552
.949
.423
.504
.705 3.0 2.5 3.0 3" Gate
.878
.876
.976
.729
.786
.814 3.0 2.5 3.0 ISX-02 42" Butterfly
.244
.436
.842
.269
.412
.609 3.0 2.5 3.0 42" Butterfly
.230
.389
.833
.238
.390
.602 3.0 2.5 3.0 42" Butterfly
.412
.799
.983
.363
.696
.758 3.0 2.5 3.0 42" Butterfly
.274
.644 1.012
.290
.554'
.738 3.0 2.5 3.0 Q110.6Sb-3
TABLE'110.66-3-continued VALVE ACCELERATIONS ON ASt'.E PIPING VALVE DESIGN BASIS SSE MARBLE HILL SSE ALLOWABLE SSE SUBSYSTEM SIZE / TYPE H
H V
H H
V H
H V
y 2
y 2
y 1SX-02 42" Butterfly
.525
.591 1.157
.454
.459
.077 3.0 2.5 3., 0 42" Butterfly
.570
.602 1.295
.559
.470
.978 3.0 2.5 3.'O 24" Butterfly
.323
.309
.991
.321
.321
.839 3.0 2.5 3.0 24" Butterfly
.79 8
.604 1.164
.699
.529 1.004 3.0 2.5 3.0 0.75" Globe
.672
.284 1.171
.630
.305
.845 3.0 2.5 3.0 24" Butterfly
.599
.396
.966
.593
.370
.714 3.0 2.5 3.0 30" Butterfly
.304 1.253
.896
.328
.937
.650 3.0 2.5 3.0 1SX-05 16" Butterfly
.518
.449 1.036
.468
.639
.875 3.0 2.5 3.0 30, 1SX-13 10" Butterfly
.768
.450 1.685
.737
.569 2.055 3.0 2.5 3.0 y[
10" Butterfly
.802
.451 1.534
.788
.568 1.899 3.0 2.5 3.0
@l 10" Butterfly
.887
.453 1.480
.890
.565 1.732 3.0 2.5 3.0
$f 10" Butterfly 1.162
.753 1.189 1.255
.796 1.149 3.0 2.5 3.0 lWO-32 3" Globe
.746
.641 1.19 1
.725
.861, 1.043 3.0 2.5 3.0 3" Globe
.850
.925 1.162
.906 1.149 1.032 3.0 2.5 3.0 3" Globe
.862 1.137 1.033
.921 1.624
.881 3.0 2.5
- 3. 0 -
3" Globe
.829 1.1669 1.324
.793 2.252 1.254 3.0 2.5 3.0 Q110,66b4
-=
l' j
TABLE 110.66-3-continued VALVE ACCELERATIONS FOR ASME CLASS PIPING 1
Valve Size Design Basis SSE Marble Hill SSE Allowable j-Subsys tem Line and Type Accelerations (G's)
Accelerations (G's)
Accelerations (G's).'
H V
H V
[RH02 MiR System 12" Gate 2.7 2.0 2.1 1.5 3.0, 2.0 (3.3J 0.9 f 3. 0 '-
2.0(()I )
)
12" Gate 4.2 0.7 12" Gate 4.6 2.6 (3.8) 1.9 (3.0 /
2.0
~
12" Gate 2.0 1.9 2.4 1.4~
3.0 2.0 15104 Accumul ator 10" Gate 3.2 1.8 3.5 1.9 4.2 3.0 System 15109 Accumulator 10" Gate 2.3 0.9 2.1 0.7 4.2 3.0 Sys tem
'E3 1CV06 CVCS 3" Gate 2.9 1.4 2.6 1.6 3.0 2.0 3" Globe 3.6 1.9 2.8 2.3 6.0 4.0 Cf ICV 60 Excess 1" Globe 0.1 0.1 0.9 0.5 3.0 2.0 n
Letdown 1" Globe 0.1 0.1 0.5 0.6 3.0 2.0 3$
1CC25 Component 2" Globe 2.3 2.1 2.8 2.4 8.5 4.0 A
Cooling 2" Globe 2.4 1.4 3.7 1.7 8.5 4.0 4" Gate 2.0 1.5 2.5 1.4 3.0 2.0 1RC21 Samole 3/4" Globe 0.1 0.0 0.7 0.5 3.0 2.0 Line (1) Higher Limits to be qualified.
0110.66b-5
TABLE 110.66-3-continued VALVE ACCELERATIONS FOR ASME CLASS PIPING valve Size Design Basis SSE Marble Hill SSE Allowable Subsystem Line and Type Accelerations (G's)
Accelerations (G's)
Accelerations (G's)..'
H V
H V
H V
1CV36 RCP Seal 3/4" Globe 2.36 3.52 2.2 3.27 8.5 4
ICC21 Component 3" Globe 1.6 1.3 0.8' O.6*
6 4
1CC22 Cooling 3" Globe 4.1 1.9 3.0 0.9 6
4 1CC23 3" Globe 4.2 2.9 1.7 1.7
,6 4
3" Globe.
- 4. 7 '
3.6 2.2 2.3 4
6(l) 3" Control 3.3 0.6 1.9 0.3 3
2 ISXO6 Service Cp Water 10" Butter-3 fly 2.4 0.38 2.42 1.31 3
2 3
10" Butter-O fly 0.62 0.38 1.65 1.16 3
2 10" Butter-fly 0.16 0.34 0.35 1.24 3
2 10" Butter-fly 0.66 0.26 1.55 1.0 3
2 ISX08-Service 10" Butterf. 1.12 0.3 1.66 1.16 3
2 Water 10" Butterf, 1.68 0.54 2.11 1.29 3
2 10" Butterf. 1.14 0.42 2.26 1.43 3
2 10" Butterf. 1.46 0.68 1.61 1.38 3
2 (1) Higher Limits to be Qualified 0110.66b-6
B/B-FSAR RESPONSE - 110.66c There are a total of 875 supports on the 40 piping problems selected.
Of these 875, only 242 supports had load increases.
Table 110.66c-1 contains the comparison of the support loads resulting from the dynamic-analysis using the design basis and Marble Hill response spectra for the SSE event.
Also included is the support number and type, and the maximum load carrying capacity of the support for service Level C.-
0110.66c-1
'rABLE 110. 66c--l
~
PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE IIILL LOAD
% INCREASE LOAD CAPACITY 1AF-02 1AF02006X 649 677 4.3 8000 1AF02016X 476 786 65.5 8000 1AF02009X 665 743 11.2 8000 1P Guido 1840 1884 2.4 1AF02013X 539 606 12.4 8000 1AF-05 1AF05061R 4748 4780
.67 6000 1AF05002X 729 762 4.5 9600 Q
lAF05058R 595 659 10,7 2250
'g*,
1AF05004R 1525 1570 2.9 2410 h
lAF05006R 588 607 3.2 1500 lAF05007R 472 496 5.1 1500 1AF05008R 406 407
. 24 1500 1AF05010 R 296 301 i.7 1500 1AF05012R 430 454 5.6 1500
- ~
lAF05014R 361 382 5.8 1500 1AF05015R 368 377 2.4 1500 1AF05017R 363 378 4.1 1500 l
1AF05018R 363 372 2.4 1500 l
lAF05020R 363 376 3.6 1500 1AF05021R 363 389 7.2 1500 1AF05023R 364 374 2.7 1500 1AF05025R 363 366
.8 1500 0110.66c-2
TABLE 110.65c-1-continued i
PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE !!ILL LOAD
% 'NCREAGE LOAD CAPACITi lAF-05 1AF05026R 376 393 4.5 1500 1AF05028R 397 422 6.3
-1500 l AF05029 R 387 408 6.4 1500 1AF11002X 3853 2871
.6 9600 1AF11003X 3023 3068 1.5 9600 1AF11021R 580 650 12.1 1500 1AF11005X 764 817 6.9 9600 lAF11006X 864 900 4.3 9600 1AF05072R 1625 1648 1.4 2410 1AF05040R 404 442 9.4 1500 (f
lAF05045R 362 421 16.3 1500 lAF05048R 361 390 8.0 1500 lAF05049R 369 392 6.2 1500 lAF05051R 377 407 7.9 1500 1AF05053R 376 403 7.2 1500 1AF05065R 380 470 23.7 870 lAF12003X 553 568 2.7 9600 lAF12002X 544 557 2.4 9600 1AF1200lX 506 2109 316.7 9600 lAF12005X 1360 1458 7.2 9600 lAF12004R 532 543 2.1 1500 1AF12006X 590 720 22.0 9600 1AF05060S NOT AVAILABLE Q110.66c-3
~
~
TABLE.110.66c-1-continued PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE IIILL LOAD
% INCREASE LOAD CAPACITY 1AF-05 1AF05063S 3936 4035 2.5 8610 1AF05064S 2428 2510 3.4 8610 lAF050093 4134 4231 2.3 8610 1AF-06 1AF0600lR 1407/-61 1415/-69 0.6 2410 1AF06013R 920/-142 937/-159 1.8 NON-STANDARD 1AF0 6017R 820 828 0.9 930 lAF06025X 736/-661 972 32 9600 lAF06026X 494/-580 618 i 4.5 9600 Q
lAF06032R 524 525 0.2 930 1AF06014R 589 627 6.5 1500
]
1AF06028X 530/-352 566 6.8 9600 lAF-07 1AF07026X 372 394 5.9 9600 1AF07009X 630 639 1.4 9600 lAF0 7016 R 729/-161 749 2.7 1500 lAF07024X 335 378 12.8 9600 R:
Rigid in Y Direction X:
Rigin in Horizontal Direction S:
Snubber G:
Guides Q110.66c-4
TABLE 110.66c-1-continued
~
PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE HILL LOAD
% 'PCREASE LOAD CAPACI'AY 1DO-16 1 dol 6004X 68/-1163 76/-1171 0.7 NON-STANDARD 1 dol 6006G 41/-20 51/-31 26.8 240 1 dol 6006G 50/-119 51/-121 1.6 3010 lDOl600 8G 44/-50 47/-53 6.0 240 3010 155/-93 155/-93 I
1DO-17 1 dol 7002X 650 714 9.8 3010 1 dol 7008G 212 224
- 5.6 6020 175 165 480 EE.
ISX-0~2 1SX02031R 32740 33045
.93 33500 1SX02035S 33906 34172
.080 50000 n
1SX-05 ISX05009X 6481 6528
.72 9600 ISX05010X 4953 5039
, 1.73 9600 1SX050llX 6007 6113 1.76 9600 l
1SX-13 ISX1300lX 3334 3758 12.7 4500 ISX13002R 3261 3632
_11.4 2710/ ROD ISX13014R 4283 4813 12.4 3770/ ROD ISXJ 3016X 2216 2250 1.5 4500 ISX13021R 1794 1905 6.2 1810/ ROD Q110.66c-5
/
TABLE 110.66c-1-continued PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE HILL LOAD
% INCREASE LOAD CAPACITY l
1SX-13 ISX13011X 1592 1629 2.3 3240 ISX13020X 3955 4140 1.7 2710/ ROD 1SX13019 R~-
1806 1869 3.4 ll30/ ROD 1SX13018X 3139 3459 10.2 4500 1SX-34 ISX3400lX 1731 1769 2.2 2250 1SX34004X 840 951 13.2 11630 1SX34007R 1437 1591 10.7 2250 1SX34012X 542 721 33.0 1500 1SX34008R 1922 2374 23.5 2710 1SX34009X 594 831 39.9 4500 k-ISX34013X 927 1210 30.5 4500 1SX34010R 3773 5144 36.3 11630 1SX34011X 1268 1714 35.2 11630 i
ISX-63 ISX6300lr 1110 1199 8.0 1500 1SX63002R 3102 3231 4.1 4500 ISX63003X 2018 2129 5.5 4500 1SX63004X 1507 1579 4.8 2250 1SX63005R 1733 1787 3.1 2710 1SX6300SX 942 945
.21 1500 1SX63007R 1671 1726 3.3 4500 1SX63008X 627 655 4.5 1500
~
0110.66c-6
~ _.
/
TABLE 110.66c-1-continued l
. PIPING SUPPORTS SUPPORT SUBSYSTEM SUPPORT DESIGN BASIS LOAD MARBLE !!ILL LOAD
% INCREASE LOAD CAPACITY 4
lWO-32 lWO32002X 1495 1531 2.4 NON-STANDARD i
lWO32993R 938 956
.95 1502 l
lWO32004X 126 134 6.3 3710 l
lWO32005R 986 994
.81 1502 lWO32006X 304 373 22.7 3710 f
lWO32008X lWO32007X 980 1001 2.1 1500 lWO32010X 326 431 32.2 NON-STANDARD l
lWO32012X 370 456 23.2 3710 lWS32013R 662 670 1.2 1502 lWO32014X 986 1034 4.8 3710 lWO32015R 572 591 3.3 870 j) lWO32016R4 327 355 8.6 3710 lWO32017R.
874 925 5.8 1500 lWO32018X 265 277 4.5 870 lWO32020X 342 361 5.5 870 lWO32023X 538 575 6.9 870 lWO32027X 1164 1170
.5 6000 l
lWO32032R-701 720 2.7 1502 lWO32038X 397 401 3.5 3710 lWS032039R 994 1043 4.9 2407 lWO32040R 573 599 4.5 NON-STANDARD lWO32042X 569 601 5.6 870 lWO32044R 1189 1221 2.7 2407 lWO32045X 223 281 26.0 6000
~
0110.66c-7 4
TABLE 110.66c-1-continued TYPE DESIGN MARBLE SYSTEM SUPPORT #
(5,R,F)
BASIS LOAD HILL LOAD CAPACITY 15104 SI4-004S S
3.1 3.8 8.61 ADD-SNUB S
8.9 9.5 NSC(New. load accepted)
SI4-009S S
9.2 9.7 13.96 ADD-Y-SN S
10.2 10.7 NSC (New load accepted)
ADD-H-SN S
4.6 5.5 NSC (New load accepted)
SI4-017S S
4.7 4.9 8.61 15I09 SI9-0205 S
9.2 10.7 70.35 SI9-021S S
6.1 7.3 8.61 ADD-2-SK R
7.2 7.7 NSC (New load accepted)
SI9-024S S
12.9 15.8 20.1 AD D-X-DS S
8.0 9.2 NSC (New load accepted)
SI9-0155 5
2.9 3.1 8.61 SI9-14 R F
5.1 5.5 NSC (New. load accepted).Q SI9-04S S
6.0 7.4 8.61 m
1CV06 1 CV06-15 R
1.1 1.2 24.84
)
ICV 06-14 S
1.2 1.4 20.1
>i 1CV06-08 5
2.0 2.3 8.61
~
1CV06-10 S
1.0 1.2 2.067 ICV 02 1CV02003 S
2.1 2.5 8.6 1.4 2.1 1
1CV02010 S
1.2 1CV02005 S
.5
.6
.6 i
1 CV02111 F
2.6 2.9 NSC (New load accepted)
Q110.66c-8
TABLE 110.66c-1-continued TYPE DESIGN MARBLE SYSTEM SUPPORT #
(S,R,F)
B ASIS LOAD _
HILL LOAD CAPACITY l RH02 RilR062R-F 25.5 26.7
'GC(New load accepted)
R!!R058S S
18.4 18.5 20.1 ADDY R
9.0 9.5 NSC(New load accepted) 1CV09 005 S
.3
.4
.9 045 S
.4
.5
.5 047 S
.2
.3
.9 01 3 R
.5
.6 2.3 014 R
.2
.3 1.2 01 6 R
.2
.3 1.5 019 R
.3
.4 2.3 020 R
.4
.5 1.3 030 S
.4
.5 2.1 024 F
.4
.6 NSC(New load accepted) 035 S
.2
.3
.5 036 R.
.4
.5
.9' h
039 R
.1
.2
.9 REI RCO2ACA F
.4
.5 NSC (New load accepted) e m
1CV60 VALVEl R
.2
.3 NSC (New load accepted)
D 1S D04 0045 5
.18
.23
.5 RE F
.25
.26 NSC (New load accepted) 15I02 ANCHOR 1 F
7.0 7.1 NSC (New load accepted)
ICC25 006R F
1.4 1.6 NSC (New load accepted) 010R F
.7 1.0 NSC (New load accepted) 013 R F
1.0 1.1 NSC ('lew load ' accepted) 016X R
2.5 2.8 6.0 017R R
1.0 1.2 1.5 009X R
.7
.9
.87 1R11-68 F
?.4 P.7 NSC (New load accepted)
IMS06 0045B S
47.2 48.1 On Hold 00lX 41.6 44.3 On Hold 0110.66c-9
~~~
TABLE 110,66c-l-continued TYPE DESIGN MARBLE SYSTEM SUPPORT #
(S,R,F)
BASIS LOAD HILL LOAD CAPACITY
- lSX08 SX08001R R
6. 9' 9.0 9.6 002R R
6.0 7.2 9.6 003R R
4.7 6.3 L 6.07 004R R
3.0 4.6 6.0 005R R
4.6 5.7 6.0 019X R
5.2 6.4 14.1-006R R
3.4 5 '. 0 NSC(New load accepted)-
1 015X R
2.9 3.6 14.1 007R1 R
4.3 5.6 NSC 008R1 R
4.8 7.0 NSC 009R1 R
4.0 6.7 NSC 010R1 R
8.9 10.5 NSC OllR1 R
5.1 10.5 NSC Ol3RL R
4.6 12.0 NSC 014R1 R
3.6 4.3 NSC 0215 0
2.3 2.8 8.6 E~
020R R
9.7 11.5 19.2
?
017S S
1.1 1.4 2.1 y
016X R
2.1 2.4 14.1 025R R
8.0
- 9. 2-19.2 7) 15003 1S0030'085 S
0.2 0.3 0.5 019R R
0.3 0.4 0.9 013S S
0.3 0.4 0.5 014X R
0.1 O.2 0.9 1CC21, 22, 23 1CC22005 R
0.9 1.1 NSC (New load accepted 023 R
0.3
.5 1.5 034 F
0.4
.5 NSC(New load accepted)
.Q110.66c-10 2
l TABLE 110.66c-1-continued l
TYPE DESIGN MARBLE SYSTEM SUPPORT #
(S,R,F)
BASIS LOAD HILL LOAD CAPACITY t
ISXO6 SX060lR R
3.2 4.3 4.5 02R R
2.4 3.8 NSC (New load accepted) 03R
.R 5.1 6.6 NSC 04R R
6.7 13.6 NSC 05R R
5.3 7.4 NSC 06R R
4.8 7.6 NSC
\\
07R R
4.8 7.9 NSC 08R R
4.3
-7.0 NSC 09R R
4.8 6.8 NSC 10R R
4.7 6.8 NSC l1R R
2.9 3.2 NSC ((New load ' accepted) 3.0 Faul ted 4.0) 14X R
3.0 4.4 ISR R
1.0 1.6 1.7 34R R
3.6 6.1 5.0(Faulted 7.1) 365 S
3.0 4.3 8.6 35X R
3.6 4.5 6.0 37X R
1.9 2.6 3.7 22R R
5.7 6.4 6.6 24X R
1.2 1.8 7.8 23X R
3.3 4.7 7.8 4
25X R
1.3 1.7 3.1 y,
27X R
.4
.6 3.7 4
26X R
2.5 2.8 3.7 76 16R R
5.3 6.4 6.6
..., 9. 0 18R R
3.4
'7'.8'-
9 0 -6,-G(Faulted 526) 17X R
3.2 5.6 14.0 19X R
1.2
- 2. 5 2.3(Faulted 3.0) 20X R
.6 1.0 2.2 21X R
2.6 3.5 6.0 28R R
4.5 6.5 6.2(Faulted 8.8) 2 9 ).
R 4.2 5.2 7.8 30X R
4.5 5.7 7.8 31X R
3.4 3.9 7.8 33X R
4.2 4.6 7.8 32X R
2.6 J.0 7.8 Q110,66c-ll'
TABLE 110.66c-1-continued TYPE DESIGN MARBLE SYSTEM SUPPORT #
(S,R,F)
BASIS LOAD HILL LOAD CAPACITY 1RY05 1RY05003 S
32.6 38.5 70.4' 008 R
21.8 22.0 37.7 010 S
17.3 18.3 20.1
'\\
1CV03 1CV03002 5
1.3 1.5 2.1 )
015 S
0.8 L O. 9' O.9/
011 R
0.9 1.1 4.5 Cp N >
Q t
T1 t;
D
~
9 Q110,66c-12 4
B/B-FSAR RESPONSE - 110.66d The difference between the results of-the' dynamic analysis using the design basis and the Marble Hill response spectra for the SSE event is" minimal.
Figures 110.66d-5 through 110.66d-8 provide a qualitative measure of the overall margin to failure for the piping problems, valves and supports based on the Marble Hill spectra.
F Q110.'66d-l
. ~
B/D-FSAR
]
FIGURE 110.66d-5
)'
' PIPING PROBLEMS' I
Q Edo Ed
=i n
.m 60 z
N i
i a
y 50 E4 a
H 1
g 1
m 40 m
mNm a
m 2
30 26
.m na m
W 20 19 17 16 e
m g
12 10 s
8 1
a4 s
Ed O
E4
> 50 50-59 60-69 70-79 80-89 90-99 D1 0
h PERCENTAGE RANGE OF ALLOWABLE UTILIZED STRESSES Z
D3 N
l 1
I i
i i
t j-j Q110.66d-2 t
L
B/B-FSAR FIGURE 110.'66d-6 VALVE ACCELERATIONS GROUP I 100
. 93 80 60 X-DIRECTION 40 20 7
h 7
8 o
> 50 50-59 60-69 70-79 80-89 90-95 cc E
100 85 3
80 E
H*
60 Y-DIRECTION m
40 8
dx 20 3
3 3
3 3
I I
I I
I I
I i
~,'
y
?> 5 0 50-59 60-69 70-79 80-89 90-95
- g 100 E4 91 no 80 N
60 b
Z-DIRECTION 8y 40 20 3
5 I
I I
I
> 50 50-59 60-69 70-79 80-89 90-95
- PERCENTAGE RANGE OF ALLOWABLE UTILIZED STRESSES Q110.66d-3
B/B-FSAR FIGURE 110.66d-7 VALVE ACCELERATIONS GROUP II 60 50 40 38 HORIZONTAL 30 23 20 19 3
15 g
10 0
3 so S
y 50 50-59 60-69 70-79 80-89
$z E
60 Q
50 m
VERTICAL 40 y
38 g
30 19 20 3
15 15 10 h
7 3
e i
I O
P 50 50-59 60-69 70-79 80-89 90-99 A
O PERCENTAGE RANGE OF ALLOWABLE UTILIZED STRESSES i
g O
em N
2 0110.66d-4
B/B-FSAR
,s-FIGURE 110.66d-8 SUPPORT LOADS 80 70
}
60 60 g
O n
.y 50 a
- =
E4y 40' e
v3 g
30 m
i
$m ka 20 a
h 8
m 11 O
10 I
9 8
6 5
2 ta N
I
> 50 50-59 60;69 70-79 80-85 90-95 PERCENTAGE RANGE OF ALLOWABLE UTILIZED STRESSES
[
0110.66d-5 L
O B/B-FSAR QUESTION 130.06 "We have reviewed your response to Question 130.6 and we conclude tnat it is not adequate and not acceptable for the following reasons:
"l)
Selection of SSE and OBE Design Earthquakes "A.. considerable portion of-your response is based on the conservatism you feel is in the SSE and OBE design
~
You also presented arguments for reducing the design earthquake to those originally proposed in the PSAR (zero period acceleration of 0.06g for OBE and 0.12 for SSE).
"These values have been subsequently increased to 0.099 and 0.2g respectively and rationale for the Regulatory staff position was stated in the Question 2.5.63.
Further-more, on the basis of further investigation, the staff came to the conclusion that the deconvolution procedures are not acceptable ant. that the Regulatory Guide 1.60
?
Design Response Spectra should be applied at the founda-tion level.
" 2)
Effect of Foundation Size on Design Spectra "The response suggests that the design spectra can be reduced based on previous studies performed by Dr. Newmark for the Diablo Canyon Site.
These studies justify reduced effective spectra as a result of considering the effect of foundation size on design spectrum.
You pointed out in the response that the reduced effective spectra were developed for the specific site of the Diablo Canyon Plant and the basic reason for its acceptance was the postulated near-field earthquake.
Since the Byron /Braidwood sites are located in an entirely different tectonic province, the argument which was used in the case of Diablo Canyon application cannot be applied to the subject sites.
"3)
Conservatism in Analysis "The staff does agree that the three components of earth-'
quake motion are probably not the same acceleration.
The magnitude of the actual acceleration of each component should be found by means of a 3-dimensional analysis.
It is the position of the staff that the response spectrum for vertical motion can be taken as 2/3 of the response spectrum for horizontal motion for the Western United i
u o
B/O-FSAR
~
States only.
For other locations, the vertical response spectrum should be the same as that given in Regulatory Guide 1.60 (see enclosure).
"As far as the damping values are-concerned, the referenced report, NUREG-CR 0098 was developed for a specific purpose of evaluating seismic risk of nuclear plants which are already operating.
The damping values contained in that report cannot be applied in licensing of new plants.
"The response claims that the elastic analysis which is used in design of new plants may be unreasonably conservative.
In view of the fact that there is a lot of safety-related equipment which might produce cata-strophic consequenes in case-of excessive deformation of supporting members, this position of the Regulatory staff is not unreasonable.
You neglected to mention in your response that the referenced criteria for the Diablo Canyon plant stipulate that the ductility of 1.3 for concrete and 3 for steel are for turbine building and intake structure.
These structures are non-Category I per se and the only reason that they have been reviewed by the staff was that in certain locations they are housing some safety-related equipment.
Thus the criteria which are applicable to these two structures cannot be automatically applied to all Category I structures.
"4)
Evaluation of Structures using 0.09g OBE and 0.20g SSE Regulatory Guide 1.60 Spectrum "The evaluation of structures using the Regulatory cri-teria provided in the response have been reviewed.
It is recognized that there is a general increase in the stress level of many structural members.
We find, however, that without re-analysis of the affected struc-tures and determination of the shear forces and moments imposed by the new loads the evaluation cannot be con-sidered to be conclusive.
You are, therefore, requested to compare the structural responses of Category I struc-tures and the design parameters (bending moments, shears and axial, loads) actually used in design of Byron /Braidwood plant with those which would have been obtained if the criteria stated 'n Question 130.06 were used."
l i
i L
RESPONSE
Introduction The design of the Byron /Braidwood structures and components required for safe shutdown was reassessed using RG 1.60 input.
This comprehensive review was made for the.SSE condition (0.20ZPA) to insure adequate safety.
The reassessment basis was agreed upon in the meeting at Bethesda on February 26, 1981 with the NRC, Commonwealth Edison Company and Sargent & Lundy.
The Marble Hill design response spectra was utilized to determine the effects on Byron /Braidwood of the RG 1.60 seismic event.
Where structural elements were identified as unique on Marble Hill, they were reassessed for Byron /Braidwood using the actual material strengths.
The Marble Hill design is based on Regulatory Guide 1.60 spectra for an SSE event of 0.20g.
The Marble Hill structures are a-replicate of the Byron /Braidwood structures.
Structural element properties as shown in Figures 130.6-1 and 130.6-2 are the same for Byron /Braidwood and Marble Hill.
For these reasons the Marble Hill design forces and spectra were used as the reassessment basis for Byron /Braidwood.
Although there are a few unique structural features on Marble Hill, the effect of these features is negligible for purposes of seismic analysis.
Comparison of the Byron /Braidwood design basis model with the Marble Hill design basis model indicates that the overall geometry mass, stiffness, and dynamic characteristics are equivalent.
The best measure of model equivalence is shown in the comparison spectra given in Appendix A.
Response spectra generated using the ' Byron /Braidwood design basis model with the same Q130.6-1
o B/O-FSAR analysis parameters'used in the Marble Hill analysis for com-parison to the Marble Hill design basis response spectra is also provided in Appendix A.
The Containment Building seismic models used for Marble Hill are identical to the models used for Byron /Braidwood stations.
Reassessment of Structures The Byron /Braidwood structures were reviewed against the Marble Hill structures by comparing the respective design drawings.
All unique elements due to the seismic design were identified.
The unique Byron /Braidwood elements were reviewed for the Marble Hill SSE loading conditions using average actual material' strengths.
For average actual material strengths, the following table of values was used:
Table 130.6-1 MATERIAL STRENGTH Design Strength Average Actual Material (psi)
Strength (psi)
Concrete 3,500 5,265 5,500 6,935 Reinfor teel (fy) 60,000 67,000*
Structut
- g. Steel (fy) 36,000 43,200*
50,000 56,000*
- Value does not exceed 70 percent of the actual average ultimate strength.
The actual concrete strengths were obtained from the concrete.
cylinder test results reported by the on-site independent testing agency.
The actual steel strengths were obtained from the certi-fied material test results submitted.by the material suppliers.
Containment Building Areas where Marble Hill had unique features were identified and a comparison of the forces was tabulated for these areas.
These areas include the base mat reinforcing, the vertical post-tension--
ing tendons, and the reinforcing steel at the main steam-penetra-tions.
i The Marble Hill and Byron /Braidwood design basis vras based on conservative assumptions as is appropriate for initial design-purposes.
0130.6-2
/*
B/B-FSAR Conservatisms were identified in the design basis analysis and the analysis was refined accordingly.
The results of the assess-ment bascJ on reanalysis show that stresses are below the design basis allowables.
Refinements to the analysi's include the following:
1.
The overall containment overturning moment, axial force and shear used in the assessment were obtained using a shell analysis rather than the beam analysis.
2.
The etiffness of the elements used to represent the reactor cavity wall was refined to account for initial concrete cracking due to shrinkage, thermal effects and restraint.
Containment Internal Structures Concrete Review of the containment internal concrete structures reveals no unique concrete elements.
Therefore, the stresses in the contaiament internal ;oncrete structures are wirhin design basis allowables.
Structural Steel Columns Review of the containment structural steel columns reveal no unique column sections Therefore, the stresses are within the design basis allowables.
4 Structural Steel Beams i
t Eighty-three beams of 740 total beams per unit have unique design due to FSE forces.
Reassessing these beams using the average actual material strength indicates a strees level below yield strength, i
Auxiliary / Fuel Handling Building i
Base Mat L
The Auxiliary / Fuel Handling Building'basemat contains 17 unique finite elements which represent less>tban 1% of the total base-mat area.
(Refer to Figure Q130.6-3 for unique areas).
Reassessi.ng these elements using the actual average material strength for the concrete and rc nforcing steel indicates a stress level below-yield strength.
Q130.6-3
B/B-FSAR Shear Walls Assessment of all shear walls based on RG 1.60 SSE spectra generated loads revealed 27 of the 65 total shear walls in the Auxiliary / Fuel Handling Building to have an increase in SSE force.
Using the average actual material strengths for the re-inforcing steel, the vertical reinforcement for all walls maintains a stress level less than the yield strength.
The
-horizontal reinforcing : steel-alsos maintained: a stress level less -
.,c than the yield strength for all walls except for one.-
This single case has a stress level of 3.5% above the yield strength.
Concrete Beams, Slabs and Piers There are no unique concrete beams, slabs, and piers between Byron /Braidwood and Marble Hill.
Therefore, all stresses are within the design basis allowables.
Structural Steel Columns of the 100 structural steel columns in the Auxiliary / Fuel Handling Building, there are 46 columns with at least one unique section.
Using the average actual material strength, the stress level did not exceed the yield strength.
Structural Steel Beams A review of the 3,400 structural steel beams in the Auxiliary /
Fuel Handling Building reveal ~ed 242 unique beams.
Reassessing these beams using the average actual material strengths indicates a stress level below yield strength.
Essential Service Cooling Tower (Byron)
The Essential Service Cooling Tower is unique to Byron Station.
The SSE forces were generated using RG 1.60 input.
Reassess-ment using these forces indicates that stress levels do not exceed the design basis allowables.
Lake Screen House (Braidwood)
The Lake Screen House is unique to the Braidwood Station.
Seismic forces were generated using RG 1.60 input.
These forces were lower than the original SSE forces used in the design and therefore all stresses renain within the design basis allowables.
Q130.6-4
~
o B/3-FSAR Electrical Raceways and Supports The assessment of raceways indicated no unique design features when compared with Marble Hill.
Raceway supports on four different elevations in the Auxiliary Building were chosen for reassessment.
These elevations have the largest increase in acceleration values in the frequency range of the raceway support.
122 hangers, representing 15
.different types, were chosenrand analyzed using Marble Hill t
response spectra.
The results of reanalysis are as follows:
Table 130.6-2 RACEWAY SUPPORTS
SUMMARY
No. of Supports No. of Supports No. of Hangers-With Ductility With Ductility Elevation Reviewed Ratio < l. 0 Ratio >l.0 477'-0" 55 49 6
451'-0" 30 30 0
401'-0" 10 10 0
426'-0" 27 27 0
The six hangers with a ductility ratio (actual stress / yield stress) greater than one have overstresses in one of their members giving a ductility ratio ranging from 1.04 to 1.33 with an average of 1.17.
These members will not collapse under an SSE event.
Conduit Supports Six conduit supports are unique to Marble Hill.
These supports were reassessed and it was determined that Byron /Braidwood supports are within the design basis allowables.
HVAC Ducts and Supports The assessment of dVAC ducts indicate no unique design features when compared with Marble Hill.
HVAC duct supports at elevation 477'-0" in the Auxiliary Building were reassessed using the Marble Hill spectra.
This elevation was chosen because it has the largest increase in acceleration Q130.6-5
O s-
'B/B-FSAR values in the frequency range of the HVAC duct' supports.
Based on a refined analysis of 347 supports, it was found that stresses do not exceed the yield strength, except for twelve cases.
These twelve supports will not collapse under an SSE event.
Piping Systems, Supports and In-line Valves Refer 1o the responses to MEB questions 110.65, 110.66,
,110.67, rl10.68 and 110.70,- Thesesquestions -comprise 'the reassess- -
a bo e ment cf the piping systems, supports and in-line valves.
Equipment Due to the different methodologies used in reassessing the safe shutdown equipment, the equipment was divided into two groups:
NSSS equipment and balance of plant equipment.
~
NSSS Equipment Introduction A reassessment was performed to determine if sufficient margin exists in the MSSS supplied mechcanical equipment to withstand a change in the Byron seismic design basis.
The investigation was performed for both primary mechanical equipment and auxiliary mechanical equipment.
A comparison was made of the generic seismic qualification levels for the equipment with'the applicable Marble Hill response spectra.
These comparisons indica te that all of the equipment has sufficient design margin to be qualified with no modification to the Marble Hil.1 response spectra.
Generic Seismic Analysis of Mechanical Equipment The NSSS supplied mechanical equipment that was included in this study is listed in Table 130.6-3.
This list contains both primary and auxiliary equipment.
All of this equipment was designed and analyzed for use in many plants.
As part of the original design, seismic qualification loads were established that-provide sufficient margin so that the equipment can be used in plants with both low and high seismic designs.
The amount of margin in the design depends on the type of analysis used to qualify the component.
Margin for Primary Mechanical Equipment Comparisons were made of the generic seismic response spectra with the Marble Hill response spectra for all of the primary equipment identified in Table 130.6-3.
An example of the comparison for the reactor vessel internals is shown in Figure Q110.70-1 This figure 0130.6-6
B/B-FSAR is typical of all the primary components comparisons and represents the component with the smallest amount of margin between the generic response spectrum and the actual plant response spectrum.
Based on these comparisons, it is clear that the response spectra in this reassessment does not significantly change the margin in the components generic design.
Margin for Auxiliary Mechanical Ecuipment Al=1 of the~ auxiliary mecha'nical equipment was qualified using ' static - ~" '
"g" levels.
For the equipment listed in Table 130.6-3, the pumps were qualified for 2.1 g's along each of the two principal horizontal axes and 2 g vertically.
The tanks and heat exchangers were qualified for 1.5 g in each of the two principal horizontal axes and 1.5 g in the vertical axes.
All of the pumps included in this study were demonstrated to be rigid by either analysis or test and an investigation of the seismic accelerations for Marble Hill spectra above 33 Hz shows the highest acceleration to be 0.99 This is well below the 2.lg acceleration used for qualification of the pumps and, therefore, all the pumps are acceptable.
The tanks and heat exchangers all have one modal frequency below 33 Hz.
A comparison was made between the qualification levels for these components and the Marble Hill seismic levels.
The com-parison was made at the calculated component natural frequency with the applicable spectra for the location of the component in the plant.
The highest seismic acceleration spectra was found to be 1.lg; this is withi.i the design acceleration for the component of 1.5g.
As a result of these comparisons, it was demonstrated that all of the auxiliery mechanical equipment have sufficient margin.
Summary A reassessment of the seismic design margin of the NSSS suppliad mechanical equipment was performed.
The investigation compared the components seismic design levels to the Marble Hill levels.
The reassessment demonstrated that all of the components have sufficient design margin.
BOP Equipment The method used to reassess the BOP safe shutdown equipment is as follows:
1.
Compare the seismic accelerations, from the appropriate Marble Hill spectra utilizing Reg. Guide 1.61 damping values against the seismic accelerations used in the existing qualificaticn report.
In some cases, the-Q130.6 ~/
B/B-FSAR seismic accelerations in the qualification reports exceeded the Marble Hill accelerations.
This result can be attributed to one or more of the following:
a.
Byron spectra exceeding the Marble Hill spectra for the equipment frequency.
b.
The vendor performed a generic qualification.
Vendor enveloped several spectra for equipment c.
located on various elevations.
2.
In cases where the Marble Hill levels exceeded the levels in the qualification report, stresses, de-flections and margins were calculated for the Marble Hill accelerations by scaling the calculated seismic stresses and deflections by the required Marble Hill to Byron acceleration ratio.
Summary Table 0130.6-4 contains the results.
Regulatory Guide 1.61 damping values were used in the reanalysis reassessment.
In addition, it should be noted that the resultant stresses were compared to the code allowable stress, not against the failure stress of the material. The results show the safe shutdown equipment can safely withstand the required seismic event.
Response to NRC Enclosure I "Information Required for Re-evaluation"
- 1. Item:
It is requested that the applicant present the infor-mation on the respective stress components due to LOCA, if applicable, and SSE and the relation of each with the specified allowable values.
Response
The LOCA load condition is only applicable to the containment structures.
Reassessment of the contain-ment structure has shown that the stresses are within the design basis allowables.
This reassessment included the combined effect of the LOCA and SSE load conditions.
0130.6-8
B/B-FSAR
- 2. Item:
Document the following:
4 The use of Marble Hill plant seismic analysis as a.
the basis for comparison with the Byron /Braidwood plant.
In your discussion, provide a comparison of the mathematical models for key structures for the two plants which show dynamic parameters such as stiffnesses, periods, moduli of elasticity, Poisson's ratio and masses.
_,.., b., Describe. any difference between -the two plants in * ' " ' "" "
terms of construction materials, quality control, construction techniques, etc.
Describe the detailed procedures as to how the c.
Marble Hill seismic responses will be used in com-puting stress levels for different load combi-nations of Byron /Braidwood structural members.
In addition, describe the criteria for selection 1
of members to be evaluated.
4
Response
a.
The Marble Hill design is based on RG 1.60 spectra for an SSE event of 0.20g.
The Marble Hill structures are a replicate of the Byron /Braidwood structures.
For these reasons, the Marble Hill design forces and-spectra were used as the reassessment basis for Byron /Braidwood.
Although there are a few unique struc-tural features on Marble Hill, the effect of these features is negligible for modeling purposes.
Comparison of the Byron /Braidwood design basis model with the Marble Hill design basis model indicates that the overall geometry, mass, stiffness, and dynamic characteristics are equivalent.
The best measure of model equivalence is documented in the comparison spectra shown in Appendix A.
Response spectra were generated using the Byron /Braidwood design basis model with the same analysis parameters used in the Marble Hill Analysis for comparison to the Marble Hill design response spectra.
For completeness, the Byron /Braidwood design basis spectra is also provided in Appendix A.
i The Containment Building seismic models used for Marble I
Hill are identical to the models used for Byron /Braidwood Stations.
b.
Equivalent material and construction specifications are used on the two plants.
c.
The Byron /Braidwood and Marble Hill structures were com-pared using their respective design drawings.
All unique elements due to seismic design were identified and reassessed using Marble Hill forces and Byron /Braidwood actual material strengths.
All design basis loading combinations are the same for Byron /Braidwood and Marble Hill.
l Q130.6-9 l
=_.
i B/B-FSAR i
- 3. Item:
Provide a comparison of the floor response spectra used in design of structures at key locations for each of the safety-related structures of the Byron /Braidwood and Marble Hill plants.
Response: A comparison of the SSE response spectra is provided in Appendix A.
- 4. Item:
In order to assess the actual margins in the design for the
,,,,,,,, OBE._ loads, compare.for-the-key member,Y the stress levels resulting from the original Byron /Braidwood seismic analysis with those resulting from the use of Marble Hill loads.
I 1
Response: Reassessment based on OBE loads was done on a few randomly selected structural elements which were reassessed earlier for increased SSE loads.
Marble Hill used a zero period acceleratbn of 0.08g OBE and Byron /Braidwood used 0.09g OBE, therefore the Marble Hill loads were factored by.09/.08 to determine the Byron /Braidwood OBE loads.
Containment Building Review of the Containment structure for Marble Hill loads factored by the ratio between Byron /Braidwood and Marble Hill
~~
OBE level indicates that the containment is within design allowable stresses.
Eight structural beams in the Containment Building were reassessed.
Using the average actual material strength, all the beam stresses are within the design basis allow-ables.
Auxiliary / Fuel Handling Building Eleven structural steel beams were selected from the Auxiliary /
Fuel Handling Building.
Using the average actual material strength, a factor of safety between 1.42 to 1.61 was maintained against yield strength.
Ten structural steel columns were reassessed for tLe Auxiliary / Fuel Handling Building.
Using the average actual material strength, a factor of safety between 2.43 to 3.35 was maintained against yield strength.
Twenty-one shear walls in the Auxiliary / Fuel Handling Building were reassessed.
Based on the average actual material strength of the reinforcing steel a factor of safety for the horizontal reinforcing steel between 1.47 to 3.45 was maintained against yield strengths.
A factor of safety in the vertical reinforcing steel between 1.61 to 6.58 was also maintained.
The stresses in these walls do not exceed the yield strength.
Q130.6-10
/*
B/B-FSAR From the finite element model of the Auxiliary / Fuel Handling l
Building basemat, ten elements were chosen in the critical SSE areas.
Using the average actual material strengths for both concrete and reinforcinq steel', the stresses of all ten elements did not exceed the yield strength.
4
- 5. Item:
Ductility under normal conditions shall be one.
Exceptions will be considered in spccial situations for SSE load and
- j. _
where modifications are cons'idered: impractical"bythe applicant and his judgement is confirmed by the staff, pro-vided that safety is assured.
Floor response spectra will be computed on the basis of elastic analysis.
Response: Ductility is the ratio between the actual stress to the yield stress.
Throughout the reassessment of the structure, a ductility ratio less than or equal to one has been main-tained.
One isolated case for a shear wall where the ductility ratio exceeded one has been indicated.
This case is the result of the SSE load condition.
Under the OBE condition, the ductility ratio does not exceed one.
Seismic responses were completed on the basis of clastic analysis.
- 6. Item:
Material Properties a.
Concrete i
As built compressive strength of concrete, f'c may be.used in the re-evaluation.
The as built strength of-concrete shall be demonstrated by the applicant through submittal of test data.
The average compressive s treng th, established by the tests, can be used as the "as built concrete strength".
The scope and the extent of tests performed shall be evaluated and approved by the staff.
b.
Steel Both reinforcing and structural steel yield stresses, fy, will be taken as the average of actual test values.
In no case will the yield strength value used in strength computations be taken as greater than 70 percent of the corresponding tested average ultimate strength value.
The scope and the extent of the test program and the resulting test data shall be reviewed and approved by the staff.
Q130.6-11
-= -.
ma. __
=
i B/B-FSAR Response: The average actual material strengths used have been given l
in Table 130.6-1.
The actual concrete strength was obtained from the concrete cylinder test results reported by the on-l site independent testing agency.
The actual steel strengths 1
were obtained from the material certification reports sub-mitted by the steel supplier.
The average actual yield strengths have been compared to the average actual ultimate 1
strengths and it was found tthat the yield does not exceed 70 percent of the ultimate.
I
- 7. Item:
Analysis Procedures a.
Regulatory Guide 1.61 damping values will be used.
b.
Accidental torsion will be considered by including an additional eccentricity in the mathematical models of 5 percent of the building dimension in the direction perpendicular to the applied loads.
c.
Stability requirements as stated in the. Standard Review l
Plan Section 3.8.5 must be met.
Response: a.
Damping values used in the reassessment conform to Regulatory Guide 1.61.
b.
The horizontal seismic model for the Auxiliary / Fuel Handling / Turbine buildings include eccentricities corresponding to the distribution of mass and stiffness for Byron /Braidwood.
The eccentricities between the mass centroid for a slab and the center of rigidity of its supporting shear walls results in an average torsional moment of 8% of the maximum building dimension times the story shear, for the major slabs in the model.
It is unlikely that these eccentricities would increase significantly due to any changes other than major changes to the plant structure since the weight of the permanent structural i elements accounts for a large percentage of the total mass.
c.
Stability requirements of the Standard Review Plan Section 3.8.5 have been met.
j Summary 1
The design of the Byron /Braidwood structures and components required for safe shutdown has been reassessed for SSE loads based on RG 1.60.
This reassessment has shown that the design of the Byron /Braidwood plant is conservative.
Byron /Braidwood seismic design basis insures that the integrity and functionality of the safety related structures is maintained.
0130.6-12
l 1
Table 130.6-3.
I I.
Primary Mechanical Equipment Reactor Pressure Vessel EN
]
Reactor Vessel Internals l
Reactor Coolant Pump Steam Generator Loop Pressurizer i
II.
Auxiliary Mechanical Equipment i
l A.
Pumps l
Centrifugal Charging Pumps RHR Pumps Boric Acid Pumps Component Cooling Pump Lube Oil Pumps for Centrifugal Charging Pump Essential Service Water Pump Motor Auxiliary Feedwater Pump Motor B.
Tanks and Heat Exchangers Boric Acid Tank i
RHR Heat Exchanger Component Cooling Heat Exchanger l
i f
i I
t
/
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raciog (Z) rato ( 2)
H1 H2 VE'T Essential Aux 1.50 1.50 1.50 3
23.9 0.42 0.42 0.44 yes Service 330' Water Pumps Aux Feed Aux 0.50 0.50 1.00 2
>33 0.38 0.36 0.65 yes Pump Motor 383' Driven Aux Feed Aux 0.50 0.50 1.00 2
733 0.38 0.36 0.65 yes Pump 383' Diesel Driven Essential Aux 0.77 0.77 0.43 2
20.9 0.33 0.33 0.32 yes Service 330' O.53 0.53 0.38 2
23.7 0.28 0.28 0.28 yes CD Water 0.48 0.48 0.36 2
24.6 0.27 0.27 0.26 yes E
Strainers 0.41 0.41 0.32 2
26.0 0.26 0.26 0.25 yes 0.20 0.20 0.20 2
35.9 0.20 0.20 0.20 yes n
L1 h
Fuel Oil Aux 0.34 0.34 0.04 2
750 0.34 0.31 0.65 yes Transfer 373' Pump & Motor 500 Gallon Aux 0.38 0.38 0.83 2
>3 3 0.47 0.37 0.65 no 87%
D/ day Tank 401 2500 Gallon Aux 0.95 0.95 0.84 4
17.5H 0.62 0.54 0.65 yes Dyf Storage 373 SRSS 3RSS 733V 0.82 0.65 Tank SRSS 500 Gallon Aux 0.38 J.38 0.83 2
>3 3 0.38 0.36 0.65 yes I
DF day Tank 383 Aux Buildina Aux 0.81
).81 0.67 2
>3 3 0.73 3.66 0.91 no 17%
Supply Fans h51' Aux Building Tux 0.96 1.96 0.67 2
73 3 0.92 3.65 0.85 no 11%
Exhaust Fans 175'
(f
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gg H
H7 VERT ESS Service Aux 2.30 2.30 2.30 2
32.7 0.21 0.21 0.21 yes Water Pump 330' Room Coolers ESS Service Aux Tested yes Pump Room Fans 330' RHR Pump Room Aux 2.30 2.30 2.30 2
32.7 0.27 0.28 0.55 yes Cubicle Coolers 346' RHR Pump Room Aux Tested yes Cubicle Fans 346' Centrifugal Aux 2.30 2.30 2.30 2
32.7 0.30 0.26 0.65 yes Charging Pump 364' m~
Room Coolers r
T Centrifugal Aux Tested yes g
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364' p
Room Fans Diesel Driven Aux 2.00 2.00 2.00 2
>33 0.38 0.36 3.65 yes Au:: Feed Pump 383' Room Cubicle Cooler Diesel Driven Aux Tested yes Aux feed Pump 383' Room Cubicle Fans ESS Switchgear Aux 0.65 0.65 0.94 2
733 0.67 0.50 0.91 no 631%
Room Fans 326' Diesel Generator Aux 0.52 0.52 0.84 2
733 0.48 3.38 0.65 yes Room Exhaust Fans 401' Misc Elect Equip Aux 0.89 0.89 0.95 2
733 0.73 3.66 0.91
.yes Room Vent Fan 451'
r T4e @pa 6"/
(CM 'd)
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fer Pump (Aux.)
Diesel Driven Aux.
Repli cate equipme lt which i s being gunericr lly qualifie-1 Aux. Feedwater 383' to 1( vels e:<ceedi ig the rec uired level.
Pump Ba tteries lAF01EA-A lAF0 LEA-B 1AF01EB-A lAF01EB-B g
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C.
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CD r
tion Bus and 451' Panels 7
l 1DC05E y
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Replicate e4 1uipment being gt;alified to enve lope o f B/B
+
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1DC01EA 451' 1DC01EB 1DC02EA 1DC02EB Battery Chargers Aux Repli cate equipment tested to envelor e of B/B + MH spectra, yes 1DC03E 451' 1DC04E Switchgear Aux.
Quali.Eicatic n reoc rts not roceived ye t.
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33 0.73 0.66 0.91 no 402%
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i Aux Feed Pump Aux Tested yes Diesel Drive 383' And Controls (EW)
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Chilled Water Aux 1.12 1.29 5.2 2%
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Refrigeration Aux 0.95g 0.95g 1.5g 2%
24.3Hz 0.70g 0.85g 1.3g yes g
N Units 383'-0 (EW)
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I Cooling Coil Aux 2.3 2.3 3.3 2%
30.2Hz 0.90g 0.67g 0.85g yes n
T.
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Cooling Tower ESWCT (0.75g, 2%
6.2Hz 1.15g no 186%
Fan Blades 0.70g 909' O.78g 2%
21.6Hz 0.64g yes Cooling Tower ESWCT Gear Box 909' l.74g 1.35g 2.04g 2%
45.9Hz 0.4g 0.29g 0.50g yes (1.5X (1.5X (1.5X 1.16) 0.90) 1.36)
Cooling Tower ESWCT Fan Motor 909' O.68g 0.68g 0.68g 2%
33Hz 0.50 0.40 0.52 yes Cooling Tower Internals Lintels ESWCT 888' O.28g 2%
13.99Hz 0.73g no
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Ta ble Q l30.6-4 Diesel Fauinment MARBLE MILL DES!CN BA5t$
SPECTRA MH $PECTRA IP NO, GIVE EQJIP=sNT LOC.
FACTOR (I)
NATU:tAL )
EN%iLOPED7 Y/it TOTAL MARGIN s>ECias DArPjNG FREQ (HZ yggy WRT Main Engine Aux 1.55 2
24.15 0.67 yes Structure 401' O.93 2
33.0 0.47 yes 0.85 2
50.58 0.62 yes 0.953 2
30.71 0.66 yes 1.047 2
30.74 0.39 yes i
0.917 2
33.35 0.38 yes 0.80 2
39.50 0.35 yes 1.025 2
27.63 0.67 yes 1.187 2
28.43 0.43 yes 0.928 2
33.10 0.38 yes 0.80 2
39.50 0.35 yes i
Fuel Oil Filter Aux 0.80 0.80 0.84 2
64.6 0.40 0.34 0.62 yes And Strainer 401' 03 Fuel Injection Aux 1.78 2.04 1.16 2
>50 1.66 1.08 0.65 yes Pump and Nozzle 401' 9
7 Fuel Transfer.
Aux 1.26 1.71 0.71 2
750 0.60 3.54 0.66 yes Q
Pump 401' p
18" Exhaust Mani-Tux 2.52 L.89 1.20 2
39.7 1.66 L.08 0.85 yes Fold Expansion 401' Joint Air Intake Aux 0.80 0.80 0.83 2
33.2 0.67 0.52 0.91 no
+43%
Filter Silencer 426' Exhaust Silencer Aux 0.80 0.80 0.83 2
37.3 0.92 0.65 0.83 no
+3.1%
477' Lube Oil Relief Aux Resu:. tant 5.0 2
750 Resu.. tant 0.85 yes Valve 401' 11= 5. O H=5, O Jacket Water Aux.
0.50 0.50 1.00 2
733 0.47 0.38 rs. 6 5 yes Circ. Pump 401'
Tahf (k l.30.6-9 D a i 'C Diesel Equipment (Cont.)
MRSLE HILL DES!CN BASIS SPECTRA MH SPECTR IF N0, Sivt NATURAL )
StECTRA DAMP!NG FREQ (HZ W LOPED NN
'OTAL m &tm EQJtP* INT LOC.
FACTOR (I)
H ERT H1 Hy VERT Jacket Water Aux.
1.20 1.18 1.94 2
750 0.85 0.76 0.74 yes Pump Engine 401' Driven Lube Oil Filters Aux.
0.80 0.80 0.85 2
750 0.47 0.38 0.65 yes i
401' Lube Oil Pump Aux.
O.89 0.89 3.18 2
Ill 7 50 0.40 0.38 0.09 yes 401' II2 7 50 0.47 0.38 1.5 yes SRSS SRSS SRSS v 717 0.62 0.54 1.64 yes Turbo Lube Oil Aux.
5.0 5.0 5.2 2
n/a 1.80 1.58 5.10 yes Filter 401' g
Jacket Water and Aux 0.80 9.30 0.05 2
)33 0.47 0.38 0.65 yes 7
Lube Oil IIeaters 401' 7,
t>
Lube Oil Circ.
Aux 0.38 0.38 0.84 2
733 0.47 0.38 0.65 no
+15%
Pump 401' I
Lube Oil Circ Aux 1.70 1.20 5.20 2
733 0.'47 0.38 0.65 yes Pump Motor 401' Starting Air Aux 3.0 3.0 3.0 2
750 OJ47 0.38 0.65 yes Relief Valve 401' Starting Air Aux 0.80 0.80 3.845 2
3 0. 8 - 111 0.50 0.38 0.65 yes Separator 401' 733-fl +V 2
Starting Air Aux 5.00 5.00 5.30 2
8. 6 - 111 0.80 1.10 3.50 yes Dryer 401' 8. 5 3 - 112 11.0-v 0.80 0.80 0.85 S 1. 9 - 111 0.40 0.34 0.64 yes 3 3. 2 - 112 37.2-v 0.80 0.80 0.85 I! 6. 3 - 111 0.40 0.34 0.62 yes 7 3. 8 - 112 112.0-v
(k l' 0. $ ~ k odI cl
& clZ Diesel Equipment (Con t. )
MARBLE HILL CESIGN BA51S
$PfCIRA M't SPECTRA IF MO, SivE N(TURAL )
gaggtopgo? y/M TOTAL MARsta
$biffRA DAp'P)qG ggggy gy g
FACTOR (I)
FREQ (NZ H]
Hy VERT I
Starting Air Aux 1.2 2
25.7 0.57 yes After Cooler 401' 1.5 24.4 0.68 yes 0.8 0.8 0.85 733 0.47 0.38 0.65 yes Jacket Water Aux 5.0 5.0 5.0 2
M3 0.53 0.48 0.89 yes Thermo Valve 401' Lube Oil Thermo Aux 5.0 5.0 5.0 2
73 3 0.63 0.42 0.65 yes Valve 401' Intercooler Aux 2.17 2.57 1.41 2
733 0.94 0.41 0.86 yes 401' 3,4 and 6" Valver Aux 3.0 3.0 3.0 2
733 0.40 0.40 0.85 yes Ball & Wafer 401' O.81 0.93 1.14 yes tt Sphere Jamesbury 0.81 0.40 0.85 yes Valves 0.46 2.16 1.03 yes 7
(n k
Gate Valves Aux 5.0 5.0 5.0 2
n/a 1.14 1.51 1.44 yes 401' Standpipe Aux 1.75 2
22.3 0.84 yes 401' l.59 23.7 0.<8 yes 0.80 0.85 45.5 0.35 yes 745 0.63 yes Fuel Oil Relief Aux 5.0 5.0 5.0 2
750
.43 0.03 0.03 yes Valve 401' Starting Air Aux 2.52 2.52 1.68 2
733 1.14 1.51 1.44 yes Compressor Motor 401' Lube Oil Cooler Aux 1.20 1.20 1.20 2
25.35 0.68 0.59 0.76 401' O.80 0.80 0.80
>33 0.48 0.38 0.65 yes
}k&c 0}11 0 b ' If (los?'0)
Diesel Equipment (Cont. )
FA"8'E "ILL cEstcN Basis SPECTRA MH SPECTR IP N0, SivE NATURAL )
sottipA DAMP!NG FREQ (HZ ENviLOPED Y/N TOTAL MARGIN EQJ!P"C1T LOC.
Factor (I) g g
gqy g}
gy yggy Support Level Switch On Aux 2
22.3 1.45 Standpipe 401' 23.7 1.12
>33 0.65 yes 3.42ZP7 1.lZPA Level
- 3. 0 2 ZP A Switch Jacket Water Aux 0.88 2
21.9 0.84 Heater System 401' O.727 23.6 0.68 Piping 0.40 55.7 0.34 0.85 750 0.65 yes Jacket Water Aux 1.60 2
15.5 0.90 Cooler Sys tem 401' 1.01 20.5 0.95 g
Piping 0.73 23.5 0.72 g
0.55 26.4 0.62 0.49 28.2 0.43 3
0.46 30.7 0.40 13 0.43 33.3 0.47 0.41 35.0 0.45 F'
O.40 36.3 0.44 no 34 Lube Oil Strainer Aux 0.68 2
24.2 0.68 Piping 401' O.40 42.6 0.4 0.40 0.85 750 0.34 0.65 yes Lube Oil Cooler Aux 1.60 2
14.9 0.90 Piping 401' O.78 23.2 0.80 0.50 27.8 0.58 0.49 28.4 0.43 0.47 29.9 0.48 0.45 31.2 0.47 0.42 34.3 0.37 0.40 36.4 0.42 0.85 740 0.65 no 40 Lube Oil Heater Aux 0.46 2
30.7 0.40 Piping 401' O.44 32.0 0.38 0.40 39.6 0.41 30 0.85 NO 0.65 no
~
sye ano.6 - V (Co.,TJ)
Diesel Equioment (Cont.)
o MA'8LE MILL DE3!CN BASIS SPECT8A MM $PECT8A IF 40, $!VE NATURAL )
EN%iLOPED7 Y/N TOTAL MAactu
?51CT*A DAMP!NG EQJtP*fNT LOC.
FACTOR (I)
FREQ (NZ
.g g y
g yg,y Lube Oil Drain Aux 0.50 0.46 0.86 2
52.7 0.43 0.37 0.66 yes Line 401' Lubs Oil Dump Aux 1.96 1.96 1.96 2
>33 0.40 0.38 0.85 yes s
Valve 401' Starting Air Aux 1.27 2
27.3 0.47 Tank Assembly 401' O.85 35.0 0.45 0.80 39.4 0.36 0.87 739.4 0.66 yes 3,5, and 6" Aux 3.0 3.0 3.0 2
250 0.40 0.40 0.85 (n.
Check valves 401' l.68 1.19 0.85 g-1.60 1.03 0.85 k.
0.40 0.40 0.85 yes tn
) Starting Air Aux 1.10 1.70 1.70 2
46 0.61 0.80 0.90 yes y
Compressor 401' 4
- Intercooler Aux 2.17 1.41 1.56 2
44.2 0.94 0.41 0.86 yes j Water Piping 401' i Jacket Water Aux 0.95 0.95 0.95 2
29.6 0.50 0.41 0.66 Cooler 401' O.87 0.87 0.87 31.4 0.49 0.39 0.65 0.80 0.80 0.80 733.1 0.47 0.38 0.65.
yes
! Generator Aux 0.80 0.80 0.85 2
750-x,y O.49 0.38 0.65 yes 401' 31,4-2 AC. Outlet Box Aux Testc d at 2% D amping.
yes 401' l Starting Air Aux 5.0 5.0 5.0 750 0.47 0.38 0.65 yes Relief Valve 401' Thermo Control Aux 0.47 1.00 1.04 34.8 2.6 14.8 22.0 yes Valve 401' L
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04 MARBLE MILL ctstcN Basts ECJIP=CNT LOC.
(!)
ha h2) 4 D Y/M OTAL R M H}
Hy vtRT N1 Hy VERT HVAC Local Aux 0.38 0.38 0.84 2%
733 0.47 0.38 0.6; no 34%
Control Panels 401' lVD0lJA lVD0lJB
' HVAC Local Aux 0.70 0.70 0.95 2%
733 0.72 0.65 0.90 no 77%
Control Panels 451' lVE01J IVX0lJ IVX02J Au::. Fe e d.
Pump Aux Re;. lica te eq uipment being generically qualified to levels k
Startup Pase' 383 exceeding the required leve:.s 1AF0lJ 9
TT bCbJ Replicate equipmen t tested t:o envelope B/B
+MH spectra.
yes 1DC11J l
Diesel Generator Aux Qualii icatior1 Repo rt not rec eived yet Control Panels 401' 1PLO7J 1PLO8J Local Instrument Cont.
Panels 377' 1PL50J IPL67J IPL75J Aux.
1P L52J 364' O.97 0.97 1.25 2%
73 3 0.40 0.36 0.75 yes APL79JB Safety IPL77JC Valve 1PL84JA Room j
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ALL STRUCTURAL ELEVENTS S
(
THElft THICKNESS SHCWN ON THIS
'e SKETCH ARE SAME FCR BYRON /
BRAIITNOOD AS WELL AS MARBLE HILL PROJECTS.
3'. 6" 3'- 6" N
t F
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l 3 ' 1 4l BED ROCK wm;;il CONTAINMENT BUILDING FIGURE 130.06-1 t
o e,
L Q
W BB AUXILIARY BUILDING FUEL HANDLLNG BUILDING e
EL 4 05'- 0" e
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NOTE:
~o ALL STRUCTURAL ELEMENTS AND-THElR THICKNESS SHOWN ON THIS SKETCH ARE SAME FOR I
B)RON/DRAIDWOOn AS WELL AS MAHDLE HILL PROJECTS.
t I
AUXtLIARY AND FUEL HANDLING BUILDINGS FIGURE 130.OG-2 s
m__.--
d' (0
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):
t e
4 k
FUEL HANDLING '
build!NG 8
c EL.342'-4"m 1
k.
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BYRON enAic 6cOc MAT PLAN AUXILIARY-FUEL HANDLING BUILDING ' COMPLEX LEGEND.
f[/////j, INDICATES UNICUE AREAS
~ ~ '
FIG URE : 130.06 -3
I B/B-FSAR o
L ADDENDUM TO RESPONSE TO QUESTION 130.0C t
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MR3LE IIILL DESIGN DASIS i
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B/B-FSAR d
E.1' 2 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO J
I PLANT STAFF (1.C.5)
POSITION:
The Operating Experience Assessment Review (CEAR) is perforced
^
by designated personnel under the Director of Nuclear Safety.
These personnel will perform the review in conjunction with others, as appropriate, to fulfill the functions li ted beloe.
It is the responsibility of the OEAR to review infor-mation from a variety of sources for operating infor=ation that may be pertinent to plant safety with feedback to plant
]
staff.
Specific responsibilities include:
1.
The OEAR function of the Office of Nuclear Safety will review infor:ation from a variety ot sources, i.e., NRC Power Reactor Events, SLA distribution letters of Bulletins, Information Notices and t
Circulars, INFO /NSAC Significant Operating Experi-4 t:nce Repor ts, etc., for operating informaticn that may be pertinent to plant safety.
This infor-nation will be transnitted in &ccordance with OEAR procedures.
Incorporaticn into training and retraining programs will be in accordance with applicable station procedures.
2.
Information that, in the judg=ent of the OEAR staff may be pertinent to plant safety, will be transmitted to plant personnel via the CEAR program.
Included will be an assessment of the inforcation and recommendations, if any, of the OEAR staff.
This inforcation, including any recon =endations, will be reviewed by the station staff in accordance with station procedures.
The station staf f will reco==end actions necessary for incorporating OEAR information.into plant procedures, operating instructions, etc.
I 3.
The OEAR will designate the applicability of the operating experience.to the appropriate organication.
For licensed plants, a station review will be performed with appropriate action i
determined (i.e.,
incorporate into operator training, review with Rad-Chem technicians).
Actions taken should be documented and fc warded to the Office of Nuclear Safety.
The OEAR will review the responses for adequacy and the uniformity of action taken throughout the company.
i f
i E.12-1 4
4 w-
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S/B-FSAR s
E.12 (Cont'd)
.t 4.
The OEAR will indicate the requirement for direct attention to the affected personnel.
Station administrative procedures will address the han-dling of OEAR's.
5.
The OEAR function of the Office of Nuclear Safety reviews information-from'a-variety of sources.
Only that viewed pertinent to plant safety will be forwarded for review by station personnel.
6.
Audits of the OEAR will be conducted by the Quality Assurance department.
l l
l l
O 4
e O
E.12-2
s o
.o p.
9/B-FSAR E.24 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION (II.D.3)
POSITION:
The reactor coolant system motor-operated pressurizer relief isolation valves (lRY8000A and B), pressurizer power-operated relief valves (lRY455A and 1RY456) and the pressurizer safety r31ief valves (lRY8010A through C) are provided with positive position indication in the main control room.
^
A discussion of the clarification items as listed in NUREG-0737 follows:
1.
The basic requirement is to provide the operator with unamb'juous indication of valve position (open or closed) so that appropriate operator actions can be taken.
For the motor-operated pressurizer relief isolation valves (lRY8000A and B), valve operator mounted limit switches provide positive position indication at the main control room valve control switch indicating lights.
For the pressurizer power operated relief valves (lRY455A and 1RY456), externally mounted limit switches provide positive position indication at the main control room valve control switch indicating lights.
For tnu pressur-izer safety relief valves (lRY8010A through C) valve s,
mounted reed switches provid2 positive position indication at the main control room valve position indicating lights.
2.
The valve position should be indicated in the control room.
An alarm should be provided in conjunction with this indication.
All valve position indications are located in the main control room.
Main control room annunciator alarms are also provided fc.- all valves.
4 3.
The valve position indication may be safety grade.
If the position indication is not safety grade, a reliable single-channel direct indication powered from a vital instrumen; bus may be provided if backup methods of determining valve position are available and are dis-i cussed in the emergency procedures as an aid to operator diagnosis of an action.
All valve position indications are safety grade.
i 1
d E.24-1 i
_ _ _. _ _, _ _. - _ - _ _. _ _ _ _ -. _~~-
_ _ _ _, - ~ _ _. _. _.
D/B-FSAR
/
4.
The valve position indication should be seismically O.'
' qualified consistent with the component or system to which it is attached.
Reference the response to clari5ication item 5 below.
J-5.
The position indic ation should be qualified for its 4
g.
appropriate environment (any transient or accident which would cause the relief or safety valve to lif t) and in accordance with Commission Order, May 23, 1980 (CLI-20-80).
For the motor-operated pressurizer relief isolation valv7s (JRY8000A and B) the position limit switches will ce ;ualified per Reference 1 of Subsection 3.11.6.
For tPc pressurizer power-operated relief valves (lRY455A and '.RY456) the position limit switches will also be qualified per Reference 1 of Substction 3.11.6.
For the pressurizer safety relief valves (lRY8010A through C) the position limit switches will be qualified to IEEE 323-1974.
All position indication in the main control room will be qualified to IEEE 323-1974.
I 6.
It is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error.
A human factor analysis should be performed taking into consid-i eration:
a.
the ure of this information by an operator during both r.ormal and abnormal plant conditions, b.
integration into emergency procedures, c.-
integration into operator training, and d.
other alarms during emergency and need for priori-tization of alarms.
The review of the main control room position indications i
for the reactor coelant system relief and safety valves is includ:d in the Byron Station Control Room Design Review for NUREG-0700.
i E.24-2
D/B-FSAR e
E.80 TRAINING AND QUALIFICATION OF OTHER PERSONNEL s
-(I.A.2.2 trom NUREG-0660)
POSITION:
For the past year Commonwealth Edison has been in the process of developing a task analysis for station operating positions, and has recently begun analyses on Radiation-Chemistry, Nuclear Engineer, Station Control Room Engineer (SCRE), and some maintenance positions.
Currently, generic task analyres have been completed for all of.the operating positions, a::d a concentrated ef fort is being made to complete the corresponding site-specific task analyses for each nuclear power plant.
Surveys, checklists, i
and interviews between task analyst and station personnel (subject matter experts) are being used by the Production Training Department to acquire the data needed for the site-specific analyses.
Each task is evaluated in terms of physi-cal difficulty,' mental difficulty, safety-related importance, operational importance, environmental conditions, and fre-quency at which the task is performed.
The task is then given a priority in relationship with other tasks within that position.
A similar approach is being taken with non-operator positions.
o.
eg_j The results of these analyses will be used eventually to establish training objectives for each of the positions.
These plans ray be altered depending on progress by INPO in the task analysis area.
\\)
E.80-1
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