ML18095A772
| ML18095A772 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/27/1991 |
| From: | Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLR-N91029, NUDOCS 9103070027 | |
| Download: ML18095A772 (5) | |
Text
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 FEB 2 7 1991 NLR-N91029 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
CYCLE 10 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit No. 1 completed its ninth cycle of operation on February 9, 1991.
The burnup at the end of Cycle 9 was 16,520 MWD/MTU.
The.startup for Cycle 10 is scheduled.for April 11, 1991.
The intent of this letter is to inform you of PSE&G's plans regarding Salem Unit No. 1 Cycle 10 reload core which is expected to achieve a burnup of 11,900 MWD/MTU.
The Cycle 10 reload core will utilize 48 new Region 12 Westinghouse 17 x 17 fuel assemblies at 4.0 w/o enrichment.
The loading contains a total of 512 fresh burnable absorber rodlets located in 40 of the Region 12 fuel assemblies, and 3,072 Integral Fuel Burnable Absorber {IFBA) rods located in all 48 of the R~gion 12 fuel assemblies as shown in the figure.
The mechanical design of the Region 12 assemblies will incorporate a modified Debris Filter Bottom Nozzle {DFBN).
The DFBN has been changed from the previous design adding a reinforcing skirt to enhance reliability during fuel handling
{Reference 1).
The IFBA rods introduced in Region 12 contain coated fuel pellets which are identical to the enriched uo2 pellets with an additional thin coating of enriched boride along the cylindrical surface of the pellet in the central portion of the fuel stack length.
In addition, the Region 12 fuel assemblies will continue to use the Reconstitutable Top Nozzle design Standardized Fuel Pellets from the previous design.
Westinghouse has completed the safety evaluation of the Cycle 10 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 f1()0 I
- I '
Document Control Desk NLR-N91029 2
FEB 2 7 199f dropped RCCA event was evaluated based on the methodology described in Reference 4.
For steam line break incidents at pressures below 1000 psia, the DNBR limit of 1.45 was utilized in the safety analysis (Reference 5).
The safety evaluation states that all Cycle 10 kinetics parameters, control rod worths, and core peaking factors meet current limits with the exception of the normalized trip reactivity insertion rate.
The normalized trip reactivity insertion rate, which is slightly different from the current limit, was compared to the previous analyses and evaluated for those accidents affected.
The analyses in the Plant Safety Evaluation (Reference 7) were demonstrated to remain applicable.
Three Salem Unit No. 1 Technical Specification changes have been approved for implementation starting in Cycle 10.
These changes are:
- 1) End of life Moderator Temperature coefficient limits changed, 2) revised heatup and cooldown curves 3) steam line pressure low setpoint changed from 500 psig to 600 psig, and steam generator low-low level changed to 16%.
The Radial Peaking Factor Limit Report for Salem Unit No. 1 Cycle 10 was previously submitted in Reference 6.
PSE&G has reviewed the basis of the Cycle 10 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
Document Control Desk NLR-N91029 3
FEB 2 7 1991 Should you have any questions regarding this transmittal, please contact us.
Sincerely, Attachment c
Mr. J. C. Stone Licensing Project Manager Mr. T. Johnson Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
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-A, July 1985.
- 3)
Salem Unit No. 1 Updated Final Safety Analysis Report, USNRC Docket Numbers 50-272 and 50-311, February 15, 1987.
- 4)
Morita, T. and Osborne, M. P., et. al., "Dropped Rod Methodology for Negative Flux Rate Trip Plants",
WCAP-10298-A, June 1983.
- 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/WCAP-9227-NP, "Reactor Core Response to Excessive Secondary Steam Releases".
- 6)
Letter from s. LaBruna to United States Nuclear Regulatory Commission, "Cycle 10 Radial *Peaking Factor Limit Report, Salem Generating Station, Unit No. 1, Docket No. 50-272",
January 29, 1991.
- 7)
Foley, J. V. and Davidson, S. L., "Plant Safety Evaluation for Salem Nuclear Plant Unit 1 and 2 Fuel Upgrade", November 1988.
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