ML18092B136
| ML18092B136 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/25/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Murley T, Varga S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), Office of Nuclear Reactor Regulation |
| References | |
| NLR-N86057, NUDOCS 8605020156 | |
| Download: ML18092B136 (4) | |
Text
j.
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 April 25, 1986 NLR-N86057 Dr. Thomas E. Murley, Administrator Region l U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Mr. Steven A. Varga, Director
- u. s. NRC PWR Project Directorate #3 Division of PWR Licensing A 7920 Norfolk Avenue Beth~~da, MD 20014 Gentlemen:
INTERPRETATION OF TECHNICAL SPECIFICATIONS FOR THE NEW WESTINGHOUSE FUEL ASSEMBLIES SALEM GENERATING STATION UNIT NOS. l AND 2 DOCKET NOS. 50-272 AND 50-311 Through discussions with Mr. D. C. Fischer, NRC Licensing Project Manager for the Salem Facility, we have become aware of differences in interpretation of a Technical Specification which has arisen at several nuclear plants that are now using new higher density fuel assemblies provided by Westinghouse Electric Corporation.
This letter clarifies the interpretation used by PSE&G in the Salem 1, Cycle 7 Reload Safety Evaluation.
The specification in question defines the maximum weight of the fuel pins to be 1766 grams.
The PSE&G interpretation has been that the maximum of the batch average pin weights in a given core reload must meet this specification.
This interpretation is based on the following facts:
- 1.
The local distribution of weight among the fuel pins within an assembly, and among assemblies, has a negligible effect on the core physics parameters and is not a significant factor in any safety analyses.
~ ~. '- -
Dr. Thomas E. Murley Mr. Steven 4/25/86
- 2.
The average pin weight in a reload batch has a very small effect on the core physics, and is considered in the design of each fuel cycle.
The attached safety evaluation, provided by Westinghouse Electric Company and reviewed by PSE&G, supports these conclusions.
The batch average fuel pin mass for region J (the new higher density fuel) is 1762 grams.
This represents a 1.0% increase from the batch H average fuel pin mass (standard fuel) which was 1745 grams.
This increase in mass has been considered in the Salem 1 Cycle 7 Reload Safety Evaluation.
The distribution of pin masses for the higher density fuel is about 6 grams, or 0.3%
at the 1 level.
This is essentially the same as the distributions of previous reloads.
The Salem 1 Cycle 7 Reload Safety Evaluation compared the Region J pin mass of 1762 grams to the Tech Spec value of 1766 and has concluded that no Tech Spec changes were required.
This same procedure was used in the Safety Evaluations of the previous Salem 1 Cycle 6, and Salem 2 Cycle 3 safety evaluations.
Based on the above interpretation, it is our position that we are, and have been, in compliance with the Technical Specifications.
However, to eliminate any future misunderstanding, we will submit a request for amendment to our licenses for Salem Unit Nos. 1 and 2 that removes that reference to individual fuel rod uranium weight.
It is our understanding that this value was deleted from the Farley Unit 2 Technical Specifications as part of Amendment No. 56 issued on April 22, 1986.
Should you have any further questions, we will be pleased to discuss them with you.
Attachment C
Mr. Donald C. Fischer Licensing Project Manager Mr. Thomas J. Kenny Senior Resident Inspector Sincerely,
SAFETY EVALUATION JUSTIFYING CONTHll ED OPERATION WITH URANIUM ROD WEIGHT DISCREPANCY The Design Features section of the Technicai Sp;cif ications identifies a maxin:n.m total weight of uraniun in each fuel rod',
Due to fuel p.ellet design improvements such as chamfered pellets with reduced dish and a nominal density increase, the fuel weight has iocreased slightly*.
The actual uraniun weight has no bearing on the p)wer limits, p::>Wer *operating level Or decay heat rate.
Al though a num~er of areas involving safety analysis are affected by fuel uran~un weight, the areas of safety significance: have their a..Jn limits which are reflected in the FSAR and Technical Sp;cif ications.
Technical
- Specifications on p:>wer and power distribution control the fission rate and, hence, the rate of decay heat production.
The corn1=0sition of the fuel is clos~ly monitqred to assure acceptable fuel perf~mance fer such things as thermal conductivity, swelling, densification, etc.
The* imp:>rtant fuel parameters have been considered and are addressed in the foll~ing evaluation as pertain.ins to Westinghouse supplied comp:>nents and services.
Seismic Effects on Fuel/Internals and Nfld and Sp;nt fuel Stora~e Racks The fuel rod uraniun weight as stated in the Technical Sp;cifications is not a direct input to the analyses of maximun seisr.iic/LOCA fuel assembly dynamic response, sei snic resp:rnse of reactor vessel and'. internals, or* seismic analyse~
of new and spent fuel storage racks *
. Radioloiical Source Terms Fission. product generation is not sensitive to ~e mass of fuel involved but to the PJWer level.
As long as the p.Jwer generated: by the core is unaffected, there will be no significant impact on the radio1.ogica1 source terms.
Fyel Handlini Aey PJStulated ircrease in the amount of uraniun in the fuel rods would not have a significant impact on the fuel handling equiµnent.
The spent fuel pit
- bridge and hoist is designed with a lced limit of approximately twice the weight of a nomiral fUel assembly.
The manipulator crane is provided with two lc.ed sensors.
One lead sensor provides primary protection of the fuel assemblies from structural damage if an assembly: were to "hang-up".
A second lc.ed sensor provides backup protection against high lift fcrce with a setpoint above that of the first lead sensor. If the setpoints were l.l'lchanged despite a slight overall iocrease in uranilJtl weight, the impact would be to decrease the potential fer fuel damage sirce reducing the difference between the fuel assembly weight and the lift fc:rce limit reduces the amount of stress the fuel assembly structure would be exp:>sed to if the as$ernbly were to "hang-upn.
The manipulator crane margin to capacity limit far exceeds any p:>tential increase in assembly weight due to irx:reases in the fuel rod uranilJtl weight.
,f LOCA Safety Ara ly sj s Ursnillil mass has no* impact on ECCS LOCA analyses. LOCA analyses are sensitive to parameters stich as pellet diameter, i:>ellet-clad gap, stack height shrinking factor and pellet density as they relate to pellet temperature and volumetric heat *generation.
Fuel rrass is not used in ECCS LOCA analyses.
non-LOCA Safety Arelysjs Individual fuel rod uraniLm weight, as rep:>rted in the Technical
- specifications, is not explicitly modeled in any* non-LOCA event.
Total uraniun present in the core is input into the transient analyses, but is generated using a methodology irxiependent of the value presented in the Technical Specifications.
Thus, any change in the number currently in the Technical Specifications does not impact the non-LOCA transient analyses.
Core Desiin The m~ss of uraniun is explicitly accounted for in the standard fuel rod design
.through appropriate modeling of the fuel i:>ellet ~eometry and initial fuel density.
Variations in uraniun mass associated with allcwable as-built variations h.lt within the specification limits r<,- the pellet dimensions and initial density are accounted for in the reactor core design analyses.
The Technical Si:;ecif ication uraniun mass value has no impact on margin to reactor core design criteria.
The conclusion of these evaluations is that there is no U'lreviewed safety question associated with operation of the U'lit(s) with a fuel rod weight in excess of that defired in Section 5.3.1 of the Technical Specifications.