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{{#Wiki_filter:*e Public Service Electric and Gas Company Stanley LaBruna                           Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President - Nuclear Operations JUL 14 1992 NLR-N92087 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
{{#Wiki_filter:Public Service Electric and Gas Company  
CYCLE 11 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem unit No. 1 completed its tenth cycle of operation on March 4, 1992. The burnup at the end of Cycle 10 was 11,818 MWD/MTU.                 The startup of Cycle 11 is scheduled for August 4, 1992. The intent of this letter is to inform you of PSE&G's plans regarding Salem Unit No. 1 Cycle 11 reload core which is expected to achieve a burnup of 15,750 MWD/MTU.
*e Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President - Nuclear Operations JUL 14 1992 NLR-N92087 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
The cycle 11 reload will utilize two regions of fresh fuel (see figure_!).                 The first region consists of two sub regions:                                   12 Region 13B assemblies enriched to 4.4 w/o u 235 each containing .48 Integral Fuel Burnable Absorber (IFBA) rods, and 16 Region 13B assemblies enriched to 4. 4 w/o                                 u2 35 containing 104 IFBA rods . .
CYCLE 11 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem unit No. 1 completed its tenth cycle of operation on March 4, 1992.
The second region. consists of 40 Region 13A assemblies enriched to 4.. o w/o u                  and containing 104 IFBA rods. The loading 235 contains a total of 368 fresh burnable absorber rodlets and 6400
The burnup at the end of Cycle 10 was 11,818 MWD/MTU.
* total IFBA rods arranged as shown in Figure 2 .
The startup of Cycle 11 is scheduled for August 4, 1992.
          .The design of the Region 13A and 13B fuel assemblies is the same as the Region 12 assemblies with the exception of an increased radius bottom end plug and an increase in length of the fuel rod.
The intent of this letter is to inform you of PSE&G's plans regarding Salem Unit No. 1 Cycle 11 reload core which is expected to achieve a burnup of 15,750 MWD/MTU.
The cycle 11 reload will utilize two regions of fresh fuel (see figure_!).
The first region consists of two sub regions:
12 Region 13B assemblies enriched to 4.4 w/o u235 each containing.48 Integral Fuel Burnable Absorber (IFBA) rods, and 16 Region 13B assemblies enriched to 4. 4 w/o u235 containing 104 IFBA rods..
The second region. consists of 40 Region 13A assemblies enriched to 4.. o w/o u235 and containing 104 IFBA rods.
The loading contains a total of 368 fresh burnable absorber rodlets and 6400
* total IFBA rods arranged as shown in Figure 2.  
. The design of the Region 13A and 13B fuel assemblies is the same as the Region 12 assemblies with the exception of an increased radius bottom end plug and an increase in length of the fuel rod.
This change was made to accommodate the shorter bottom end plug.
This change was made to accommodate the shorter bottom end plug.
As a result of the fuel rod length change, the plenum spring length was increased to accommodate_ the plenum length change.
As a result of the fuel rod length change, the plenum spring length was increased to accommodate_ the plenum length change.
Since these changes do not affect normal plant operating parameters,* safeguard systems, or assumptions used in the safety analysis, these changes do not compromise the performance of any safety related system, nor result in any adverse etfect on the safety analysis.
Since these changes do not affect normal plant operating parameters,* safeguard systems, or assumptions used in the safety analysis, these changes do not compromise the performance of any safety related system, nor result in any adverse etfect on the safety analysis.
* The IFBA coated fuel pellets which were introduced in Region 12 for Cycle 10 are identical to the enriched uo pellets except for 2
The IFBA coated fuel pellets which were introduced in Region 12 for Cycle 10 are identical to the enriched uo2 pellets except for 9207210019 6~86b~72  
9207210019 6~86b~72
~DR ADOCK PDR  
        ~DR         ADOCK             PDR


Document Control Desk           2             JUL 14 1992 NLR-N92087 the addition of a thin enriched ZrB coating- on_ the pellet cylindrical surface along the centr~l portion of the stack length.   -                **                  -
Document Control Desk NLR-N92087 2
JUL 14 1992 the addition of a thin enriched ZrB coating-on_ the pellet cylindrical surface along the centr~l portion of the stack length.
Westinghouse has completed the safety*evaluation of the Cycle 11 -
Westinghouse has completed the safety*evaluation of the Cycle 11 -
reload core design utilizing the methodology described in Reference 1. Based on this methodology, those incidents analyzed
reload core design utilizing the methodology described in Reference 1.
.and reported in *the Salem UFSAR (Reference 2) that could     *
Based on this methodology, those incidents analyzed  
- potentially be affected by the fuel reload are addressed. The dropped RCCA incidents were conservatively-evaluated assuming the.
.and reported in *the Salem UFSAR (Reference 2) that could  
removal of the Negative Flux Rate Trip (NFRT) function based on the approved methodology_ in Reference 3. For steam line break incidents at pressures below 1000 psia, the DNBR limit of 1.45 -
- potentially be affected by the fuel reload are addressed.
The dropped RCCA incidents were conservatively-evaluated assuming the.
removal of the Negative Flux Rate Trip (NFRT) function based on the approved methodology_ in Reference 3.
For steam line break incidents at pressures below 1000 psia, the DNBR limit of 1.45 -
was-utilized in the safety analysis :(Reference 4).
was-utilized in the safety analysis :(Reference 4).
* The Salem Unit 1 Cycle 11 core contains a Region- 10 assembly (K-24 in location C-08) which has a single stainless steel filler rod as a result of fuel reconstitution to replace a defective fuel rod which was found by ultrasonic testing during the 8th refueling outage. The cause of the defect was determined to be debris induced fretting (Reference 6). The reconstituted fuel assembly has a predicted peak rod power _about 8% lower when compared to the core limit F-[)elta-H of 1.55. The therma,1-hydraulic evaluation for fuel rod reconstitution has been performed by Westinghouse in accordance with approved NRC codes and methods (References 7 and 8). This evaluation has assessed the safety signiflcance of the* fuel reconstitution and has
The Salem Unit 1 Cycle 11 core contains a Region-10 assembly (K-24 in location C-08) which has a single stainless steel filler rod as a result of fuel reconstitution to replace a defective fuel rod which was found by ultrasonic testing during the 8th refueling outage.
The cause of the defect was determined to be debris induced fretting (Reference 6).
The reconstituted fuel assembly has a predicted peak rod power _about 8% lower when compared to the core limit F-[)elta-H of 1.55.
The therma,1-hydraulic evaluation for fuel rod reconstitution has been performed by Westinghouse in accordance with approved NRC codes and methods (References 7 and 8).
This evaluation has assessed the safety signiflcance of the* fuel reconstitution and has
_assured that a core with a stainless steel filler rod meets the design criteria for the existing fuel design (Reference 9).
_assured that a core with a stainless steel filler rod meets the design criteria for the existing fuel design (Reference 9).
Large Break LOCA analyses.have been traditionally performed using a symmetric, chopped cosine axial power shape. _Recent calculations have shown that there was a potential for top-skewed
Large Break LOCA analyses.have been traditionally performed using a symmetric, chopped cosine axial power shape. _Recent calculations have shown that there was a potential for top-skewed  
'power distributions to result in peak claddin*g temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution. Westinghouse has developed a process, which was applied to the reload for Salem Unit 1 Cycle 11, that reasonable ensures that the cosine remains the limiting power distribution, by-defining-appropriate power distribution surveillance data.
'power distributions to result in peak claddin*g temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution.
Westinghouse has developed a process, which was applied to the reload for Salem Unit 1 Cycle 11, that reasonable ensures that the cosine remains the limiting power distribution, by-defining-appropriate power distribution surveillance data.
This process, called the Power Shape sensitivity Model (PSSM), is described in topical report WCAP-12909-P.
This process, called the Power Shape sensitivity Model (PSSM), is described in topical report WCAP-12909-P.
The auxiliary feedwater flow and the containment spray* delay issues identified by LERs 91-036-00 and 92-002-0b have been addressed for S~lem Unit 1 Cycle 11.
The auxiliary feedwater flow and the containment spray* delay issues identified by LERs 91-036-00 and 92-002-0b have been addressed for S~lem Unit 1 Cycle 11.  


*~
*~
Document Control Desk             3               JUL 14 1992 NLR-N92087 The safety evaluation states that all Cycle 11 kinetics parameters, control rod worths, and core peaking factors meet the current limits with the exception of the normalized trip reactivity insertion rate which is slightly different from the current limit.
Document Control Desk NLR-N92087 3
* The normalized trip reactivity insertion rate was compared to the previous analyses and evaluated for those*
JUL 14 1992 The safety evaluation states that all Cycle 11 kinetics parameters, control rod worths, and core peaking factors meet the current limits with the exception of the normalized trip reactivity insertion rate which is slightly different from the current limit.
    .accidents affected. The analyses in the Salem UFSAR (Reference
* The normalized trip reactivity insertion rate was compared to the previous analyses and evaluated for those*  
.accidents affected.
The analyses in the Salem UFSAR (Reference
: 2) were demonstrated to remain applicable.
: 2) were demonstrated to remain applicable.
A review of the Salem Unit 1 Cycle 11 Reload Safety         _
A review of the Salem Unit 1 Cycle 11 Reload Safety Evaluation (RSE) has been performed relative to.the impact of this RSE on the Salem Unit 1 Technical Specifications.
Evaluation (RSE) has been performed relative to.the impact of this RSE on the Salem Unit 1 Technical Specifications. A~ a result of this review,* no Technical Specification changes are required based on the subject RSE for Cycle 11 operations..
A~ a result of this review,* no Technical Specification changes are required based on the subject RSE for Cycle 11 operations..
* The Radial.Peaking Factor Limit Report for Salem Unit No. 1 Cycle 11 was previously submitted in References.*
* The Radial.Peaking Factor Limit Report for Salem Unit No. 1 Cycle 11 was previously submitted in References.*
PSE&G has* reviewed the basis of the cycle 1*1 reload analysis and the Westinghouse Reload Safety Evaluation Report with
PSE&G has* reviewed the basis of the cycle 1*1 reload analysis and the Westinghouse Reload Safety Evaluation Report with
* Westinghouse. We have determined .-that all the postulated events are within allowable limits and that no unreviewed safety questions as defined by 10CFR50.59 are involved with this reload.
* Westinghouse.
The reload core design will be verified during the startup physics testing program . . The program will include, but is not limited to the following tests:
We have determined.-that all the postulated events are within allowable limits and that no unreviewed safety questions as defined by 10CFR50.59 are involved with this reload.
: 1. control rod .drive tests and drop time measurements
The reload core design will be verified during the startup physics testing program.. The program will include, but is not limited to the following tests:
: 2. Critical boron concentration measurements
: 1.
: 3. Control rod bank worth measurements
control rod.drive tests and drop time measurements
: 4. Moderator temperature coefficient measurements
: 2.
: 5. Power distribution measurements using the incore flux mapping system
Critical boron concentration measurements
    *should you have any questions regarding this transmittal' please contact.us.
: 3.
Sincerely, Attachments
Control rod bank worth measurements
: 4.
Moderator temperature coefficient measurements
: 5.
Power distribution measurements using the incore flux mapping system  
*should you have any questions regarding this transmittal' please contact.us.
Sincerely, Attachments  


..                                                  JUL 1 4 1992 Document Control Desk             4 NLR~N92087 c   Mr. T. T. Martin, Administrator - Region I U. s. Nuclear Regulatory Commission' 475 Allendale Road King of Prussia, PA 19406
Document Control Desk NLR~N92087 4
* Mr. J. C. S.tone, Licensing Project Manager - Salem
c Mr. T. T. Martin, Administrator - Region I U. s. Nuclear Regulatory Commission' 475 Allendale Road King of Prussia, PA 19406
        .U. S. Nuclear Regulatory Commission One White. Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. T. P. Johnson (809)
* JUL 1 4 1992 Mr. J. C. S.tone, Licensing Project Manager -
USNRC Senior Resident Inspector Mr. K. Tosch, Chief NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
Salem  
.U. S. Nuclear Regulatory Commission One White. Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. T. P. Johnson (809)
USNRC Senior Resident Inspector Mr. K. Tosch, Chief NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625  


JUL 1 4 1992
. "' 0 SALEM UNIT 1 CYCLE 11 REDESIGN R
."' 0 SALEM UNIT 1 CYCLE 11 REDESIGN
I I
                                                                      . MAY 1992 @.
I !
FIGURE 1 REVISED CORE LOADING PAITERN
I I I
* SALEM UNIT 1 - CYCLE 11 R      p  N     M   L   K     J   H F'
I 11A 138 12 138 12 138 11 A p
E D      c  B  A I
118 12 12 12
I
. 138 11A 138 12 12 12 10 LEGEND FIGURE 1 REVISED CORE LOADING PAITERN
                                      '    I I
* SALEM UNIT 1 - CYCLE 11 N
            !I I
M L
I I I      11A 13B   12 13B   12   13B 11A         I j            1 I; .      10   12   12   12 138 11A   13B   12 12 12    118            2 118 11A   13A   12 13A   11A 10   1-1 A 13A 12 13A   ~nA  10        3 12 13A  11A 13A   7   13A 10   13A   7 13A 11A    13A 12        4 11A    12  12   13A   BB 13A   118 138   118   13A BB 13A    12  12 11A  - 5 138    12 13A   7   13A 11B   12 11A   12   11B 13A 7    13A  12 138  - 6_____,,.. ./
K J
12 . 138  11A   13A 118   12   118 138   118   12 118 13A    10 138  12  - 7 138    11A  10   10 138 11A   138 118   138  11A 138 10    10 11A 138  - 8 12    138 11A   iJA 118   12   118 138   118   12 118 13A   11A 138  12  - 9 138      12 13A    7   13A   10   12 11A   12   11B 13A 7    13A  12 138  - 10 11 A    12  12  -13A 88 13A   118 138   118   13A BB 13A     12  12 11A  - 11 12  13A  11A 13A   7   13A 10   13A   7 13A 11A    13A 12        12 10  11A   13A   12 13A   11A 10   11A   13A 12 13A    11A 118        13 118 ,-12   12 12   138 11A   138   12 12 12     10             14 11A 138     12 138   12   138 11A                       15 LEGEND REGION IDENTIFIER
H F'
E I I I
11A 13B 12 13B 12 13B 11A 10 12 12 12 138 11A 13B 12 12 11A 13A 12 13A 11A 10 1-1 A 13A 12 13A 11A 13A 7
13A 10 13A 7
13A 12 13A BB 13A 118 138 118 13A BB 13A 7
13A 11B 12 11A 12 11B 13A 11A 13A 118 12 118 138 118 12 118 10 10 138 11A 138 118 138 11A 138 11A iJA 118 12 118 138 118 12 118 13A 7
13A 10 12 11A 12 11B 13A 12  
-13A 88 13A 118 138 118 13A BB 13A 11A 13A 7
13A 10 13A 7
13A 11A 13A 12 13A 11A 10 11A 13A 12 118  
,-12 12 12 138 11A 138 12 12 11A 138 12 138 12 138 11A REGION IDENTIFIER JUL 1 4 1992
. MAY 1992 @.
D c
B A
I j
1 12 118 2
13A
~nA 10 3
11A 13A 12 4
13A 12 12 11A 5
7 13A 12 138 6
./
13A 10 138 12 7
10 10 11A 138 8
13A 11A 138 12 9
7 13A 12 138 10 13A 12 12 11A -
11 11A 13A 12 12 13A 11A 118 13 12 10 14 15  


9                     JUL~ 4 1992 0   SALEM UNIT 1 CYCLE 11 REDESIGN                                                 MAY 1992   @
9 JUL~ 4 1992  
FIGURE 2 BURNABLE ABSORBER AND SOURCE. ROD LOCATIONS SALEM UNIT 1 - CYCLE 11 R     P     N     Y       ~   I(
-0 SALEM UNIT 1 CYCLE 11 REDESIGN MAY 1992 @
FIGURE 2 BURNABLE ABSORBER AND SOURCE. ROD LOCATIONS SALEM UNIT 1 - CYCLE 11 R
P N
Y  
~
I(
I J
I J
I HI C
I H
F I
C I
E I
F E
D       C     8   A 481         481       481 II            -  1 8P
I I
* SP                           ~
D C
1041       1041
8 A
                                                                                                -2 I                            12P                     12P                   .
I I
1041                  . 4SSA                   1041                     -3 II                      SP 1041 4P         4P 1041 8P I        1041                                                              1041             -  4.
I I I 481 SP 1041 481 8P 1041 481 481 481 481 I
1041       1041       1041'     1041 '
8P
8P           4P         8P         4P           8P 1041       1041         1041       1041         1041
* SP  
                                                                                                  - ~
~
12P           4P                               4P             "12P 481  ,                                                                                  481 1041         1041                               1041             1041 SP            4P                       12P                     4P               8P
1041 1041 12P 12P 1041 1041
                                                                                                  - 7 1041          1041                    1041                   1041             1041 8P         12P         12P       8P 481                                                                                      481  - 8 1041       1041       1041       1041 8P            4P                       12P                     4P               8P
. 4SSA 1041 1041 SP 4P 4P 8P 1041 1041 1041 1041' 1041 '
                '                                                                                  - g 1041          1041                    1041                   1041             1041 12P            4P                               4P               12P 481                      1041                              1041             1041 481  - 10 1041 8P 1041 4P 1041 1041 4P 1041 1041
1041 8P 4P 8P 4P 8P 1041 1041 1041 1041 1041 12P 4P 4P "12P 1041 1041 1041 1041 12P 1041 1041 4P 12P 4P 1041 1041 1041 8P 12P 12P 8P 1041 1041 1041 1041 4P 12P 4P 1041 1041 1041 4P 4P 12P 1041 1041 1041 8P 4P 4P 1041 1041 1041 1041 1041 IP 4P 4P 8P 1041 1041 1041 1041 1041 1041 12P 4SSA 12P 1041 1041 1041 IP IP 1041 1041 481  
                                                                                                  - 11 IP         4P         4P         8P 1041                                                             1041                 12 1041       1041       1041       1041 12P                     12P 1041                     4SSA                    1041                          13 1041                   1041 IP         IP 14 1041       1041 481         .481       481                                         15 00 TYPE                                                       TOTAL f#P .** (MJMBER OF' PYREX ROD LETS) ***** ~ * * * * * * *
.481 481 00 TYPE TOTAL f#P.** (MJMBER OF' PYREX ROD LETS) ***** ~ * * * * * * *
* 3A f#fl ** (NUUBER OF' IF'BA RODS) *********** ~****** l400 fSSA ** (tlJMBER OF SECON>ARY SOLRa: ROOLETS)...                 I}}
* 3A f#fl ** (NUUBER OF' IF'BA RODS) *********** ~****** l400 fSSA ** (tlJMBER OF SECON>ARY SOLRa: ROOLETS)...
I 481 8P 1041 481 8P 1041 481 1
-2
-3
: 4.
~
- 7 8
g 10 11 12 13 14 15}}

Latest revision as of 02:22, 6 January 2025

Informs NRC of Plans Re Salem Unit 1 Cycle 11 Reload Core Scheduled for 920804.Cycle 11 Reload Core Expected to Achieve Burnup of 15,750 Mwd/Mtu.Encl Figure Reflects Revised Core Loading Pattern for Unit 1 Cycle 11
ML18096A837
Person / Time
Site: Salem PSEG icon.png
Issue date: 07/14/1992
From: Labruna S
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLR-N92087, NUDOCS 9207210019
Download: ML18096A837 (6)
v • d • e


Text

Public Service Electric and Gas Company

  • e Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President - Nuclear Operations JUL 14 1992 NLR-N92087 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

CYCLE 11 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem unit No. 1 completed its tenth cycle of operation on March 4, 1992.

The burnup at the end of Cycle 10 was 11,818 MWD/MTU.

The startup of Cycle 11 is scheduled for August 4, 1992.

The intent of this letter is to inform you of PSE&G's plans regarding Salem Unit No. 1 Cycle 11 reload core which is expected to achieve a burnup of 15,750 MWD/MTU.

The cycle 11 reload will utilize two regions of fresh fuel (see figure_!).

The first region consists of two sub regions:

12 Region 13B assemblies enriched to 4.4 w/o u235 each containing.48 Integral Fuel Burnable Absorber (IFBA) rods, and 16 Region 13B assemblies enriched to 4. 4 w/o u235 containing 104 IFBA rods..

The second region. consists of 40 Region 13A assemblies enriched to 4.. o w/o u235 and containing 104 IFBA rods.

The loading contains a total of 368 fresh burnable absorber rodlets and 6400

  • total IFBA rods arranged as shown in Figure 2.

. The design of the Region 13A and 13B fuel assemblies is the same as the Region 12 assemblies with the exception of an increased radius bottom end plug and an increase in length of the fuel rod.

This change was made to accommodate the shorter bottom end plug.

As a result of the fuel rod length change, the plenum spring length was increased to accommodate_ the plenum length change.

Since these changes do not affect normal plant operating parameters,* safeguard systems, or assumptions used in the safety analysis, these changes do not compromise the performance of any safety related system, nor result in any adverse etfect on the safety analysis.

The IFBA coated fuel pellets which were introduced in Region 12 for Cycle 10 are identical to the enriched uo2 pellets except for 9207210019 6~86b~72

~DR ADOCK PDR

Document Control Desk NLR-N92087 2

JUL 14 1992 the addition of a thin enriched ZrB coating-on_ the pellet cylindrical surface along the centr~l portion of the stack length.

Westinghouse has completed the safety*evaluation of the Cycle 11 -

reload core design utilizing the methodology described in Reference 1.

Based on this methodology, those incidents analyzed

.and reported in *the Salem UFSAR (Reference 2) that could

- potentially be affected by the fuel reload are addressed.

The dropped RCCA incidents were conservatively-evaluated assuming the.

removal of the Negative Flux Rate Trip (NFRT) function based on the approved methodology_ in Reference 3.

For steam line break incidents at pressures below 1000 psia, the DNBR limit of 1.45 -

was-utilized in the safety analysis :(Reference 4).

The Salem Unit 1 Cycle 11 core contains a Region-10 assembly (K-24 in location C-08) which has a single stainless steel filler rod as a result of fuel reconstitution to replace a defective fuel rod which was found by ultrasonic testing during the 8th refueling outage.

The cause of the defect was determined to be debris induced fretting (Reference 6).

The reconstituted fuel assembly has a predicted peak rod power _about 8% lower when compared to the core limit F-[)elta-H of 1.55.

The therma,1-hydraulic evaluation for fuel rod reconstitution has been performed by Westinghouse in accordance with approved NRC codes and methods (References 7 and 8).

This evaluation has assessed the safety signiflcance of the* fuel reconstitution and has

_assured that a core with a stainless steel filler rod meets the design criteria for the existing fuel design (Reference 9).

Large Break LOCA analyses.have been traditionally performed using a symmetric, chopped cosine axial power shape. _Recent calculations have shown that there was a potential for top-skewed

'power distributions to result in peak claddin*g temperatures (PCT) greater than those calculated with a chopped cosine axial power distribution.

Westinghouse has developed a process, which was applied to the reload for Salem Unit 1 Cycle 11, that reasonable ensures that the cosine remains the limiting power distribution, by-defining-appropriate power distribution surveillance data.

This process, called the Power Shape sensitivity Model (PSSM), is described in topical report WCAP-12909-P.

The auxiliary feedwater flow and the containment spray* delay issues identified by LERs 91-036-00 and 92-002-0b have been addressed for S~lem Unit 1 Cycle 11.

  • ~

Document Control Desk NLR-N92087 3

JUL 14 1992 The safety evaluation states that all Cycle 11 kinetics parameters, control rod worths, and core peaking factors meet the current limits with the exception of the normalized trip reactivity insertion rate which is slightly different from the current limit.

  • The normalized trip reactivity insertion rate was compared to the previous analyses and evaluated for those*

.accidents affected.

The analyses in the Salem UFSAR (Reference

2) were demonstrated to remain applicable.

A review of the Salem Unit 1 Cycle 11 Reload Safety Evaluation (RSE) has been performed relative to.the impact of this RSE on the Salem Unit 1 Technical Specifications.

A~ a result of this review,* no Technical Specification changes are required based on the subject RSE for Cycle 11 operations..

  • The Radial.Peaking Factor Limit Report for Salem Unit No. 1 Cycle 11 was previously submitted in References.*

PSE&G has* reviewed the basis of the cycle 1*1 reload analysis and the Westinghouse Reload Safety Evaluation Report with

We have determined.-that all the postulated events are within allowable limits and that no unreviewed safety questions as defined by 10CFR50.59 are involved with this reload.

The reload core design will be verified during the startup physics testing program.. The program will include, but is not limited to the following tests:

1.

control rod.drive tests and drop time measurements

2.

Critical boron concentration measurements

3.

Control rod bank worth measurements

4.

Moderator temperature coefficient measurements

5.

Power distribution measurements using the incore flux mapping system

  • should you have any questions regarding this transmittal' please contact.us.

Sincerely, Attachments

Document Control Desk NLR~N92087 4

c Mr. T. T. Martin, Administrator - Region I U. s. Nuclear Regulatory Commission' 475 Allendale Road King of Prussia, PA 19406

  • JUL 1 4 1992 Mr. J. C. S.tone, Licensing Project Manager -

Salem

.U. S. Nuclear Regulatory Commission One White. Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. T. P. Johnson (809)

USNRC Senior Resident Inspector Mr. K. Tosch, Chief NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625

. "' 0 SALEM UNIT 1 CYCLE 11 REDESIGN R

I I

I !

I I I

I 11A 138 12 138 12 138 11 A p

118 12 12 12

. 138 11A 138 12 12 12 10 LEGEND FIGURE 1 REVISED CORE LOADING PAITERN

  • SALEM UNIT 1 - CYCLE 11 N

M L

K J

H F'

E I I I

11A 13B 12 13B 12 13B 11A 10 12 12 12 138 11A 13B 12 12 11A 13A 12 13A 11A 10 1-1 A 13A 12 13A 11A 13A 7

13A 10 13A 7

13A 12 13A BB 13A 118 138 118 13A BB 13A 7

13A 11B 12 11A 12 11B 13A 11A 13A 118 12 118 138 118 12 118 10 10 138 11A 138 118 138 11A 138 11A iJA 118 12 118 138 118 12 118 13A 7

13A 10 12 11A 12 11B 13A 12

-13A 88 13A 118 138 118 13A BB 13A 11A 13A 7

13A 10 13A 7

13A 11A 13A 12 13A 11A 10 11A 13A 12 118

,-12 12 12 138 11A 138 12 12 11A 138 12 138 12 138 11A REGION IDENTIFIER JUL 1 4 1992

. MAY 1992 @.

D c

B A

I j

1 12 118 2

13A

~nA 10 3

11A 13A 12 4

13A 12 12 11A 5

7 13A 12 138 6

./

13A 10 138 12 7

10 10 11A 138 8

13A 11A 138 12 9

7 13A 12 138 10 13A 12 12 11A -

11 11A 13A 12 12 13A 11A 118 13 12 10 14 15

9 JUL~ 4 1992

-0 SALEM UNIT 1 CYCLE 11 REDESIGN MAY 1992 @

FIGURE 2 BURNABLE ABSORBER AND SOURCE. ROD LOCATIONS SALEM UNIT 1 - CYCLE 11 R

P N

Y

~

I(

I J

I H

C I

F E

I I

D C

8 A

I I

I I I 481 SP 1041 481 8P 1041 481 481 481 481 I

8P

~

1041 1041 12P 12P 1041 1041

. 4SSA 1041 1041 SP 4P 4P 8P 1041 1041 1041 1041' 1041 '

1041 8P 4P 8P 4P 8P 1041 1041 1041 1041 1041 12P 4P 4P "12P 1041 1041 1041 1041 12P 1041 1041 4P 12P 4P 1041 1041 1041 8P 12P 12P 8P 1041 1041 1041 1041 4P 12P 4P 1041 1041 1041 4P 4P 12P 1041 1041 1041 8P 4P 4P 1041 1041 1041 1041 1041 IP 4P 4P 8P 1041 1041 1041 1041 1041 1041 12P 4SSA 12P 1041 1041 1041 IP IP 1041 1041 481

.481 481 00 TYPE TOTAL f#P.** (MJMBER OF' PYREX ROD LETS) ***** ~ * * * * * * *

  • 3A f#fl ** (NUUBER OF' IF'BA RODS) *********** ~****** l400 fSSA ** (tlJMBER OF SECON>ARY SOLRa: ROOLETS)...

I 481 8P 1041 481 8P 1041 481 1

-2

-3

4.

~

- 7 8

g 10 11 12 13 14 15

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