ML18092B052
| ML18092B052 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/12/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NLR-N86028, NUDOCS 8603210106 | |
| Download: ML18092B052 (4) | |
Text
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Public Service Electric and Gas Company Corbin A. McNeill, Jr.
Vice President -
Public Service Electric and Gas Company P.O. Box236, Han cocks Bridge, NJ 08038 609 339-4800 Nuclear March 12, 1986 NLR-N86028 Off ice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20014 Attention:
Mr. Steven A. Varga, Director PWR Project Directorate #3 Division of PWR Licensing A Gentlemen:
CYCLE 7 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit No. 1 is scheduled to complete its sixth cycle of operation on March 21, 1986.
The burnup at the end of Cycle 6 is predicted to be approximately 17,100 MWD/MTU.
The startup of Cycle 7 is scheduled for May 21, 1986.
The intent of this letter is to inform you of PSE&G's plans regarding Salem Unit No. 1 Cycle 7 reload core which is expected to achieve a burnup of 16,000 MWD/MTU.
The Cycle 7 reload core will utiliz*e 84 new Region 9 Westinghouse 17xl7 fuel assemblies at 3.8 w/o enrichment, and 1616 fresh burnable poison rods.
The mechanical design of the Region 9 fuel assemblies is the same as the Region 8 assemblies, except for the use of chamfered fuel pellets and 304L stainless steel grid sleeve material.
The fuel pellets have a small chamfer at the outer edge of the pellet ends and a reduction in the dish diameter and depth compared to previous unchamfered pellets.
Westinghouse indicates that the chamfer will improve chip resistance during manufacturing and handling, and that the new grid sleeve material will further reduce the potential for stress corrosion cracking of the grid sleeves.
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Mr. Steven 3-12-86 Westinghouse has completed the safety evaluation of the Cycle 7 reload core design in accordance with the Westinghouse reload methodology as outlined in the Westinghouse topical report,"Westinghouse Reload Safety Evaluation Methodology, 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.
Several Unit 1 Technical Specification changes have been proposed for implementation starting in Cycle 7.
These changes consist of a power uprate to 3411 MWt from 3338 MWt, and the removal of the boron injection tank (BIT).
The power uprate has been approved by the NRC, while the BIT removal is awaiting NRC approval.
Failure to attain the BIT removal will not alter the conclusions determined in the Westinghouse Reload Safety Evaluation.
This evaluation states that all Cycle 7 peaking factors, rod worths, and kinetics parameter values meet current limits, and that no safety reanalysis of any incidents was necessary.
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.
PSE&G has reviewed the bases of the Cycle 7 reload analysis and the Westinghouse Reload Safety Evaluation Report with Westinghouse.
In addition, PSE&G has performed an independent reload safety evaluation for Cycle 7 using in-house computer codes.
In both cases it was demonstrated that the results of all the postulated events are within allowable limits.
We have determined 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 Salem Unit 1 operating license is not required.
The reload core design will be verified during the startup physics testing program.
This 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
.)
I/
- Mr. Steven A.Varga 3-12-86
- 4.
Moderator temperature coefficient measurement
- 5.
Power distribution measurements using the incore flux mapping system.
Should you have any questions, we will be pleased to discuss them with you.
Attachments
References:
Sincerely,
- 1.
Morita, T., *osborne, M.P., et ~l., "Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A (proprietary} and WCAP-10298-A (Non-Proprietary}, June 1983.
C Mr. Donald C. Fischer Licensing Project Manager Mr. Thomas J. Kenny Senior Resident Inspector
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