ML18096A837

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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)
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Text

  • 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:

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 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 . .

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

  • 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 uo pellets except for 2

9207210019 6~86b~72

~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. - ** -

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 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.

  • 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

  • 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.

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

.. 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

  • 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

JUL 1 4 1992

."' 0 SALEM UNIT 1 CYCLE 11 REDESIGN

. MAY 1992 @.

FIGURE 1 REVISED CORE LOADING PAITERN

  • SALEM UNIT 1 - CYCLE 11 R p N M L K J H F'

E D c B A I

I

' I I

!I I

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_____,,.. ./

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

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 HI C

F I

E I

D C 8 A 481 481 481 II - 1 8P

1041 1041

-2 I 12P 12P .

1041 . 4SSA 1041 -3 II SP 1041 4P 4P 1041 8P I 1041 1041 - 4.

1041 1041 1041' 1041 '

8P 4P 8P 4P 8P 1041 1041 1041 1041 1041

- ~

12P 4P 4P "12P 481 , 481 1041 1041 1041 1041 SP 4P 12P 4P 8P

- 7 1041 1041 1041 1041 1041 8P 12P 12P 8P 481 481 - 8 1041 1041 1041 1041 8P 4P 12P 4P 8P

' - 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

- 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) ***** ~ * * * * * * *

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