ML20196D439
| ML20196D439 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/19/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20196D436 | List: |
| References | |
| NUDOCS 8802170142 | |
| Download: ML20196D439 (5) | |
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SAFETY EVALUATION SY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 38 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT N0. 30 TO FACILITY OPERATING LICENSE NPF-52 DUKE POWER COMPANY, ET AL.
DOCKET NOS. 50-413 AND 50-414 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 INTRODUCTION By letter dated September 8, 1987, Duke Power Company, et al., (the licensee) proposed amendments to revise Technical Specification 5.3.1 "Fuel Assemblies" by increasing the maximum allowable fuel enrichment to 4.0 weignt percent (w/o)
U-235 from the present value of 3.5 w/o U-?35 EVALUATION The principal safety considerations associated with the proposed amendments are the potential effect of using the mere highly enriched fuel in the reactor core and the criticality aspects of storing 4.0 w/o U-235 fuel assemblies as new fuel and as spent fuel.
Before any of the fuel enriched above 3.5 w/o U-235 is loaded into the reactor core, its higher enrichment will be included in the cycle-specific reload safety evaluation (RSEi, which considers in detail the effect of fuel enrichment on core operating parameters.
The RSE will use the standard reload design methods described in the Topical Reports WCAP-9272 and 9273, "Westinghouse Reload Safety Evaluation Methodology." The proposed amendments are, therefore, acceptable with regard to core reload because the use of fuel enriched up to 4.0 w/o U-235 will be expressly taken into account in the final safety evaluation of each cycle-specific core reload.
Criticality accidents during refueling operations are precluded by stringent administrative procedures.
The criticality analyses for new and spent fuel storage described in Sections 9.1.1 and 9.2.2 of the Final Safety Analysis Report are based on a fuel enrichment of 3.5 w/o U-235. A summary of the criticality analyses for the same storage facilities but for the increased fuel enrichment is nrovided by the licensee in A of its September 8,1987, submittal.
For the new fuel storage vault and for the spent fuel pool these analyses present the criticality design criteria, a description of the facility, the methods used for the analysis, data j
on the benchmarking of the analysis methods, and the criticality analysis results.
The design basis for preventing criticality in both fuel storage facilities is taken from ANSI N18.2-1973, Section 5.7.4.1, which states:
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"The design of spent fuel storage racks and transfer equipment shall he such that the effective multiplication factor will not exceed 0.95 with new fuel of the h'ighest anticipated enrichment in place assuming flooding with pure water. The design of normally dry new fuel storage racks shall be such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderation."
For the spent fuel pool the accidents considered include:
- 1) loss of spent fuel pool cooling, 2) the sliding of free standing rack modules such that peripheral cells of two rack modules have center-to-center spacings below those assumed in normal design basis analyses, and 3) the dropping of fuel assemblies on top of a rack module or lowering of a fuel assembly by the side of a rack module in a non-storage location.
The criticality analysis method used for both storage facilities makes use of computer codes CSAS2 and 123 GROUPMTP for the determination of cross-sections and for the calculation of the effective multiplication factor, k Calcu-lations using this code set were benchmarked against a set of 40 8Ntical experiments representing a diverse group of water moderated oxide fuel arrays separated by materials such as stainless steel, Boral, water, etc. The comparisons indicated that there is a 95 percent probability at the 95 percent confidence level that the uncertainty in reactivity, due to the methods, is not greater than 0.012 4 k.
For the new storage vault, a number of criticality) analyses considering a full loadina of either Westinghouse 17x17 Standard (STD or Optimized (0FA) fuel assemblies were perfortneo using aqueous moderator densitics ranging from 0.05 to 1.0 en/cc.
The following assumptions were used in the criticality evaluation:
- 1) Nominal values for the fuel assembly parameters.
- 2) Credit is taken for the inherent neutron absorption in full length structural materials as allowed by ANSI N18.2-1973.
- 3) No burnable poisons, control rods, or supplemental neutron poisons are assumed to be present.
- 4) Effects of reflectors other than water are included if their neglect would have been nonconservative.
This includes the storage vault's concrete walls, ceiling, and floor.
- 5) All assemblies are assumed to be 4.1 w/o U-235 enriched and unirradiated.
This worst-case enrichment assumption allows for a specified maximum nominal enrichment of 4.0 w/o U-235 with an enrichment tolerance of 0.1 w/o U-235,
- 6) The new fuel storage vault is conservatively modeled as an infinite series of
? infinite rows of 12 foot high fuel assemblies in minimal thickness SS304 cell enclosures.
7)
Each fuel assembly is treated es a heterogeneous system with the fuel pins, control rod guide tubes, and instrumentation thimble guide tube modeled explicitly.
- 8) Fechanical uncertainties and biases due to construction tolerances are con-sidered by using worst-case conditions.
Uncertainties censidered include cell I.D., center-to-center spacing, and cell enclosure thickness.
The results of the analysis, with due allowance for calculational uncertainty and bias show an effective multiplication factor less than 0.95 for either the pure water or the optimum acueous foam condition.
Based on our review, we agree with the results of the licensee's criticality analysis and therefore find that storage of fuel with the maximum enrichment of 4.0 w/o U-235 permitted by Tech-nical Specification E.3.1 in the new fuel storage vault is acceptable.
Similar criticality analyses were performed for storing)either spent fuel or new fuel (Westinghouse 17x17 Standard or Optimized (0FA ) enriched to 4.0 w/o 11-235 in the spent fuel pocl.
The following assumptions were used in the criticality evaluatinn:
- 1) Nomir,al values for the fuel assembly parameters.
71 Credit is taken for the inherent neutron absorption in full lencth structural materials as allowed by ANSI N18.2-1973.
- 3) No burnable poisons, control roos, or supplemental neutron poisons are assumed to be present.
41 All assembifes are assumed to be unirradiated 4.05 w/o U-235 enriched Westingheuse STD or 0FA type.
This worst case assumption allows for a specified nximum nominal enrichment of 4.0 w/o U-235 with an enrichment tolerance of 0.0E w/o U-235.
- 5) The < pent fuel storage array is conservatively modeled as infinite in lateral and axial extent.
- 6) Geometrical and material uncertainties que to mechanical tolerances are treated by either usino worst-case configuration or by performing sensitivity calculations and cbtaining appropriate uncertainty values.
The uncertainties considered include:
Fuel enrichment Water density Stainless steel cell wall thickness Center-to-center spacing Cell ID Cell bowing Assembly positioning
- 7) Each fuel assembly is treated as a heterogeneous system with the fuel pins, control rod guide tubes, and instrument guide tube modeled explicitly.
- 8) The moderator is pure, unborated full density water.
. The calculated "worst-case" value of the effective multiplication facter for the Westinghouse 17x17 STD fuel is 0.9377; for the Westinghcuse 17x17 0FA fuel, k
is 0.9455 For both types of fuel the maximum k is les BI((donourreview,weagreewiththeresultsofthe*Ncensee'sthan0.95.
s analysis and, therefore, find that the storage of fuel with the maximum enrichment of 4.0 w/o U-235 permitted by Technical Specification 5.3.1 in the spent fuel pool is acceptable.
ENVIRONMENTAL CONSIDERATION These amendments involve changes to the use of facility components located with-in the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occu-pational exposures. The NRC staff has made a determination that the amendments involve no significant hazards consideration, and there has been no public comment on such finding. Accordirgly, the amendments meet the eli criteria for categ)orical exclusion set forth in 10 CFR 51.22(c Pursuant to 10 CFR 51.22(b no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
CONCLUSION The Commission made a proposed cetermination that the amendments involve no significant hazards consideration which was published in the Federal Reaister (52 FR 47783) on December 16, 1987.
The Commission consulted with the state of South Carolina.
No public comments were received, and the state of South Carolina did not have any coments.
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of these altendments will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributors:
S. Stanley Kirslis, PDII-3/DRPI/II K. ilabbour, PDII-3/DRPI/II Dated:
January 19, 1988
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DATED: January 19, 1988 AMENDMENT N0. 38 TO FACILITY OPERATING LICENSE NPF Catawba Nuclear Station, Unit 1 AMENDMENT NO. 30 TO FACILITY OPERATING LICENSE NPF Catawba Nuclear Station, Unit 2 DISTRIBUTION:
f DOCketaFile 50 413/414 NRC PDR Local PDR PRC System NSIC PDII-3 Reading M. Rood K. Jabbour S. Kirslis OGC-Bethesda D. Hagan J. Stone E. Jordan L. Harmon W. Jones T. Barnhart (8)
ACRS (10)
GPA/PA ARM /LFMB L. Reyes S. Varga/G. Lainas E. Butcher W. Hodges i
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