ML18093A540

Informs of Util Plans Re Plant Cycle 8 Reload Core,Expected to Achieve Burnup of 16,000 Mwd/Mtu.Amend to License Not Required Based on Review of Cycle 8 Reload Analysis & Westinghouse Reload SER

ML18093A540
Person / Time
Site:	Salem
Issue date:	12/16/1987
From:	Corbin McNeil Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NLR-N87227, NUDOCS 8712280185
Download: ML18093A540 (4)

Text

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Public Service Electric and Gas Company

- Corbin A. McNeil!, Jr. Public Service Electric and Gas Company P.O. Box 236, Han cocks Bridge, NJ 08038 609 339-4800 Senicr Vice President -

Nuclear December 16, 1987 NLR-N87227 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

CYCLE 8 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 SALEM GENERATING STATION UNIT NO. 1 DOGKET NO. 50-272 Salem Unit No. 1 has completed its seventh cycle of operation on October 2, 1987. The burnup at the end of Cycle 7 was 16,804 MWD/MTU. The startup of Cycle 8 is scheduled for December 29,

1987. The intent of this letter is to inform you of PSE&G' s plans regarding Salem Unit No. 1 Cycle 8 reload core which is expected to achieve a burnup of 16,000 MWD/MTU.

The Cycle 8 reload core will utilize 84 new Region 10 Westinghouse 17 X 17 fuel assemblies of 3.8 w/o-enrichment, and 1,712 fresh burnable poison rods. The mechanical design of the Region 10 fuel assemblies is the same as the Region 9 assemblies, except for the use of the Reconstitutable Top Nozzle (RTN), a changed bottom end plug design, and the 4g pellet holddown spring. The RTN features a removable nozzle which facilitates removal of fuel rods for examination and repl~cement. The bottom end plug is about .18 inches longer than that previously used and facilitates removal and insertion of fuel rods. The 4g spring provides more fuel gas plenum volume for fission gas release while still providing appropriate hold down force to counteract maximum anticipated handling loads.

Westinghouse has completed the safety evaluation of the Cycle 8 reload core design in accordance with the Westinghouse topical report, "Westinghouse Reload Safety Evaluation Methodoloqy WCAP-9273-A, July 1985". Based on this methodology, those incidents analyzed and reported in the Salem UFSAR that could potentially be affected by the fuel reload were addressed.

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nncument Control Desk 2 12-16-87 Three Salem Unit 1 Technical Specification changes have been approved for implementation starting in Cycle 8. These changes are {l) removal of the boron injection tank {BIT), {2) introduction of an Fxy surveillance peaking factor limit report, and (3) increasing the minimum boron concentrations of the refueling water storage tank (RWST) and accumulators.

The Reload Safety Evaluation states that all Cycle 8 peaking factors, rod worths, and kinetics parameters meet current safety limits. The* dropped RCCA event was analyzed according to the dropped rod methodology described in Reference 1. Results show that the DNB design basis is met for all dropped rod events initiated .from full power. For steam line break incidents at pressures below 1,000 psia, the DNBR limit of 1.45 was utilized in the safety analysis (Reference 2).

PSE&G has reviewed the basis of the Cycle 8 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. Therefore, based on this review, application for amendment to the Sal~m Unit No. 1 operating license is not required.

The Radial Peaking Factor Limit Report for Salem Unit 1 Cycle 8 was submitted previously in Reference 3.

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

1. Control rod drive tests and drop time

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

Document Control Desk 3 12-16-87 Should you have any questions, please do not hesitate to contact us.

Sincerely,

References:

1) Mori ta T ;, Osbourne, M. P., et al., "Dropped Rod Methodology for Negative Flux Rate Trip Plants,"

WCAP-10297-P-A (proprietary), and WCAP-10298-A (non-proprietary), June 1983

2) Westinghouse letter dated March 25, 1986, NS-NRC-86-3116, "Westinghouse Response to

• Additional Request on WCAP-9226-P/WCAP""."9227-N-P, "Reactor Core Response to Excessive Secondary System Release", (non-proprietary)

3) Letter from c. A. McNeil!, Jr. (PSE&G) to NRC, "Cycle 8 Radial Peaking Factor Limit Report, Salem Generating Station, Unit No. 1, Docket No. 50-272", October 13, 1987 C Mr. D. C. Fischer USNRC Licensing Project Manager Mr. T. J. Kenny USNRC Senior Resident Inspector Mr. w. T. Russell, Administrator USNRC Region I Mr. D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628

e " e ATTACHMENT SALSM f.JNIT 1, CYCLE 8 CORE LOADING PATTERN R p N M. L (( J H G F E 0 c B ..

9 10 s 10 9 10 9 12 7* 10 10 10.

2 9

12 I10 9

2A 9

2*

9 10 12

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7* 9 10 9 9 8A* 9 10 9 9 7* I 3 20 2A SS 2A 110 20 l I 9 10 20 9 10

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9 10 24 9 10 24 9

4 10 24 I 9 110 20 I 10 9

i 9 10 9 10 9 10 a 10 e 10 9

• I 10 9 9

!I 12 2A a 2A 2A a 2* 2* 12 10 9 10 9 10 9 10 S* 10 9 10 9 I 10 9 e 110 2A 4 24 24 2* 2* 24 1

9 10 9 10 a 10 9 10 9 10 I 10' 9 10 g 2* 2A 24 2* 2A 24 24 a

10 9 IA* 9 10 6* 10 , ,. 10 6* 10 9 IA* 9 10 12 2* 2* 2* 2* 12 9

9 10 2A 9 10 2A 6 10 2A 9 10 24 9 10 2*

6 10 2*

9 10 .

2A I 9 10 9 10 10 10 e* 10 9 10 9 10 9 10 10 9

2*

9

ZA 2* . 24 10' 2* 2*

9 10 9 9 10 I 10 5 10 9 10- 9 10 9 11 12 2A 8 2A 24 2* I 24 12 9 10 9 10 9 10 9 10 9 10 9 10 9 12 7*

20 10 2A 10

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12 REG IOU W/O U235

• FROH CYCLE 6 6 3.40 ~ REGION NUMBl!ll 7 3.40 NUMBER OF BURNAS~ ASSORSER ROOS 8A 3.40 9 3.80 SS SECONDARY SOURCE LOCATIONS 10 3.SO

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