ML18096A729

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Advises That Facility Completed Sixth Cycle of Operation on 911109.Cycle 7 Reload Core Expected to Achieve Burnup of 14,000 Mwd/Mtu.Cycle 7 Reload Will Utilize Two Regions of Fresh Fuel.Startup Physics Tests Listed
ML18096A729
Person / Time
Site: Salem 
Issue date: 05/15/1992
From: Labruna S
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLR-N92063, NUDOCS 9205270135
Download: ML18096A729 (7)
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Text

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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 MAY l 5 1992.

NLR-N92063 United States Nuclear Requlatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

CYCLE 7 RELOAD ANALYSIS SALEM GENERATING STATION UNIT NO. 2

  • FACILITY OPERATING LICENSE DPR-75 DOCKET NO. 50-311 Salem Unit No. 2 has completed its sixth cycle of operation on November 9, 1991. The burnup at the end of Cycle 6 was 15,994 MWD/MTU.

The intent of this *letter is to inform you of Salem Unit No. 2 Cycle 7 reload core which is expected to achieve a burnup of 14,000 MWD/MTU.

The Cycle 7 reload will utilize two regions of fresh fuel (See Figure 1). The first consists of two sub regions: 16 Region* 9A assemblies enriched to 3.8 w/o and 8 Region 9B assemblies enriched to 4.0 w/o. The second region consists of 28 Region 12 assemblies enriched to 4.0 w/o. The loading contains a total of 544 fresh burnable absorber rodlets and 2560 Integral Fuel Burnable Absorber (IFBA) rods arranged as shown in Figure 2. The mechanical design of the Region 9 fuel assemblies will incorporate a modified Debris Filter Bottom Nozzle (DFBN) and increased radius bottom end plug.

The mechanical design of the Region 12 fuel assemblies is identical to the Region 9 fuel assemblies with one exception. The'Region 12 fuel assemblies did not incorporate the increased radius bottom end plug.

The (DFBN) has been modified by adding a reinforcing skirt to enhance reliability during postulated adverse handling conditions while refueling (Ref. 1). The bottom end plug has an increased radius in the transition between the chamfer and the end of the plug. There are no changes in the critical dimensions of the bottom end plug* or the pressure drop from the previous region.

Therefore the fuel rod performance and the core safety considerations are not adversely affected. The IFBA coated fuel pellets introduced in Regions 9 and 12 are identical to the.

enriched uranium dioxide pellets except for the addition of a thin enriched zirconium diboride coating on the pellet fD\\

cylindrical surface along the central portion of the stack

~

length.

  • )ti lfr, {1&!.

l I 1 \\ -

Document Control Desk NLR-N92063 2

MAY 15 1992.

Westinghouse has completed the safety evaluation of Cycle *7

  • reload core design utilizing the methodology described in Reference 2.

Based on this methodology, those incidents analyzed and reported in the Salem UFSAR (Reference 3) 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 dropped rod methodology (Reference 4). However, the evaluation for the dropped RCCA incidents with Negative Flux Rate Trip (NFRT) function, based on approved methodology, is also bounded. For steam line break incidents at pressures below 1000 psia, the DNBR limit of 1. 45 was utilized in the safety analysis (Reference 5).

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 cladding 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 2 Cycle 7, that 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 a topical report (WCAP-12935).

The auxiliary f eedwater flow and the containment spray delay issues identified by LERs 50-272/91-036-00 and 50-272/92-002-00 have been addressed for Salem Unit 2 Cycle 7.

That is, the*

safety analyses account for AFW flows greater than originally assumed for the steam line break analyses; and an increase to the total containment spray system delay time.

The safety evaluation states that all Cycle 7 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 Sal~m UFSAR (Reference 3) were demonstrated to remain applicable.

A review of the Salem Unit 2, Cycle 7 Reload Safety Evaluation (RSE) has been performed relative to the impact of this RSE on the Salem Unit 2 Technical Specifications (Reference 3). As a result of this review, no Technical Specification changes are required based on the subject RSE for cycle 7 operation.

The Radial Peaking Factor Limit Report for Salem Unit No. 2 Cycle 7 was previously submitted in Reference 6.

Document Control Desk NLR-N92063 3

MAY l 5 1992 PSE&G has reviewed the basis of the cycle 7 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 10CFRS0.59 are involved with this reload.

Therefore, based on this review, application for amendment to the Salem Unit No. 2 operating license is not required.

The reload core design is verified during the startup physics testing program. The program includes, 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

s.

Power distribution measurements using the incore flux mapping system Should you have any questions, we will be pleased to discuss them with you.

Sincerely,

Document Control Desk NLR-N92063 4

MAY 1 s 1992 References :

1) Letter D. w. Perone (Westinghouse) to R. J. Gennone (PSE&G), "Modified Bottom Nozzle", October 24, 1989.
2) Davidson, s. L. (Ed.), et. al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9273-NP-A, July 1985.
3) Salem Units 1 and 2 Updated Final Safety Analysis Report, USNRC Docket Numbers 50-272 and 50-311, July 22, 1989.
4) Haessler, R. L., et. al., "Me~hodology for the Analysis of the Dropped Rod Event," WCAP-11394-A, January 1990.
5) Letter from A. c. Thadani (NRC) to W. J. Johnson (Westinghouse), January 31, 1989,

Subject:

Acceptance for Referencing of Licensing* Topical Report WCAP-9226-P/92.27-NP, "Reactor Core Response to Excessive Secondary Steam Releases."

6) Letter from S. LaBruna to United States* Nuclear Regulatory Commission, "Cycle 7 Radial Peaking Factor Limit Report, Salem Generating Station, Unit No. 2, Docket No. 50-311," January 27, 1992.

/

~-

1 __ -~-----

Document Control Desk NLR-N92063 5

c Mr. T. T. Martin, Administrator - Region I

u. s. Nuclear Regulatory Commission 475 Allendale Road King.of Prussia, PA 19406 Mr. J. c. Stone, Licensing Project Manager - Salem
u. s. Nuclear Regulatory Commission One White Fl.int North 11555 Rockvil!'e Pike Rockville, MD 20852 Mr. T. P. Johnson (S09)

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 MAY 1 5 1992

R p

N M

I I

I I

I T36 U68 I

K-2 8-7 T38 U22 V21 P-6 A-8 FEED U47 V1B T56 J-14 FEED K-8 U24 XM01 U19 XM1B f.t-5 FEED L-6 FEED U41 U37 U10 R3B L-14 N-6 K-7 D-2 U61 XM63 T40 VOS D-9 FEED R-7 FEED XM06 uoa T70 POB FEED J-8 C-3 E-5 U70 XM29 T39 V01 D-7 FEED R-9 FEED U39 U67 U36 R56 L-2 N-10 K-9 D-14 U12 XM08 U27 XM33 M-11 FEED L-10 FEED U51 V20 T69 J-2 FEED H-10 T32 U03 V22 P-10 H-1 FEED T26 U58 K-14 8-9

  • ASSE
    • ASSE MBLY FROM CYCLE 5

MBLY FROM CYCLE 4

      • ASSEMBLY FROM CYCLE 3

DREGION 4 (3.410*/0 )

DREGION SA (3.808*/0 )

DREGION II (3.7118*/0 )

DREGION 7A (4.003*/0 )

._..,,n.. 'm&""*+/-,:e:-e:r* "'-

FIGURE 1 e

SALEM UNIT 2 CYCLE 7 CORE LOADING PA'ITERN L

K J

H G

F E

D U11 U53 USS XM11 U66 USO U14 L-4 8-5 G-12 FEED J-12 P-5 E-4 XM02 U38 XM54 U13 XM48 U73 XM09 U49 FEED K-3 FEED H-7 FEED F-3 FEED P-7 U25 U20 T12 TB2 T24 U16 U33 V23 K-5 J-6 J-1 C-13 G-1 G-6 F-5 FEED XM22 R02 V14 P17 V11 R23 XM49 T61 FEED P-12 FEED E-11 FEED 8-12 FEED H-6 T 14 US9 T09 U02 T02 U57 T10 XM71 C-11 N-4 L-1 H-5 K-15 C-4 L-13 FEED USS P63 V13 T45 V09 P24 U44 R53 M-3 E-13 FEED D-14 FEED N-11 D-3 M-2 T01 V1S us T06 T16 V16 T33 V04 A-10 FEED L-15 D-2 R-5 FEED R-10 FEED U09 T04 T13 R15 T25 T19 U28 P70 L-8 8-4 P-4 J-: 10 8-12 p.-..12 E-8 L-11 T42 VOS us T30 T43 V07 U7 VOS A-6 FEED A-11 M-14 E-1 FEED R-6 FEED U71 P48 V02 T21 V10 P64 U74 R32 M-1-3 C-5 FEED M-2 FEED L-3 D-13 M-14 T23 U52 T 11 U15 U.4 U69 S16 XM25 E-3 N-12 F-1 H-11 K-1 C-12 M-4 FEED XM61 R59 V03 P57 V12 R12 XM36 T7B FEED P-4 FEED L-5 FEED 8-4 FEED F-8 U26 U07 T17 T51 T03 U32 U29 V17 K-11 J-10 J-15 N-3 G-15 G-10 F-11 FEED XM03 U75 XM23 U21 XM40 U40 XM10 U62 FEED K-13 FEED H-9 FEED F-13 FEED P-9 U35 U43 U56 XM12 U60 U72 U04 L-12 8-11 G-4 FEED J-4 P-11 E:...12 oo DREGION 7B (3.903*/0 )

DREGION DREGION SA (3.1112*/0 )

DREGION DREGION 118 (3. m* 1 o>

llA (3.llO!*/o)

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B A

T34 F-2 U34 T07 H-15 8-6 V24 U42 FEED G-14 U17 XM04 U06 E-6 FEED D-5 U01 U63 U48 F-7 C-6 E-14 T1B XM2B U76 A-7 FEED M-9 T71 U30 XM07 N-13 G-8 FEED T1S XM30 U46 A-9 FEED M-7 U18 U4S U6.4 F-9 C-10 E-2 U05 XMOS U31 E-10 FEED D-11 V19 U54 FEED G-2 U23 T28 R-B 8-10 T35 F-14 98 (4.002*/0 )

12 (4.001*/0 )

DREGION

[::::J IEST I NGHOUSE ASSEJIBL Y I 0 PREVIOUS CYCLE LOCA Tl ON Figure 2-3 Salem Unit 2, Cycle 7 Randomized Core Loading Pattern Core Description 2-11 2

3 4

5 6

7 8

9 10 11 12 13 14 15

IESTIMC>IOOSE

~IETARY C~ASS 2 F'IGURE 2 SALEM UNIT 2 (0 NJ) CYCLE 7 CORE COMPONENTS AND FRESH IFBA LOCATIONS p

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I 12P I

RCCA 12P I 12P I RCCA 64! RCCA 641 RCCA I

8P I

641 RCCA RCCA 4SSAiRCCA 8P.

i2p 12P 12P RCCA 64! RCCA RCCA 641 64!

641 I 12P RCCA 12p 641 RCCA 12P 12P RCCA 641 RCCA 64[

12P 12P 12P 64[ RCCA 641 641 RCCA RCCA RCCA RCCA 12P 12P 12P 641 RCCA 641 641 RCCA RCCA 12P 641 RCCA 12P 641 12P 12P RCCA 541 8P 12?

12P 12P RCCA 541 RCCA 541 641 RCCA 641 8P 541 RCCA RCCA 4SSA RCCA 12P 12P RCCA 12P RCCA 641 RCCA 641 LE GENO

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TYPE COMPONENT TYPE l!!!:J ###1 NUMBER OF FRESH IFBA ROOS COM;~ma el - CdNTMOL di RifbOIW eel

  1. SSA - ~

r# ~LITS ON S£CON)MY SOUltCI: ASSDll. T f#' - ~

rl al.ITS I H ft'nllD ASSEll9~ Y 3 of 18 RCCA 12P 64[

RCCA 12P 641 RCCA RCCA E

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12P IRCCA 2

8P RCCA 641 12P 8P 641 RCCA 641 f~CCA 12P 641 RCCA 12P 5

RCCA 5

12P

~2P RCCA 641 64[

7 RCCA RCCA 3

i 12P 12P i

641 RCCA 641 i

3 I

RCCA

- *o 12P 641 RCCA i2p 12P BP 641 RCCA 641 RCCA SP RCCA 641 12P RCCA 15 FUEL ASSEMBLY ORIENTATION

  • RE~RENCE HOLE ocg I' IN 1-iOLE

/ H L OOIN BAR NOTE: ALL FIGURES ARE TOP v1::*t SEE TABLE 2 FOR COWPCNE~T ORIENTATION

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