ML20210V183
| ML20210V183 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/09/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210V096 | List: |
| References | |
| NUDOCS 8702180708 | |
| Download: ML20210V183 (5) | |
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k UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION 5
j WASHINGTON, D. C. 20655
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 92 TO FACILITY OPERATING LICENSE N0. OPR-36 MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309
1.0 INTRODUCTION
By letter (Ref.1) dated November 25, 1986, the Maine Yankee Atomic Power Company (MYAPCo or the licensee) made application to revise the Technical Specifications appended to Facility Operating License No. DPR-36 for operation of the Maine Yankee plant. The proposed change would increase the fuel enrichment limit of the reactor core in order to pennit increased cycle lengths. The current Technical Specifications permit a 3.3 weight percent uranium-235 fuel enrichment limit. The proposed change would increase the fuel enrichment limit to 3.5 weight percent uranium-235.
The staff has reviewed the proposed change and prepared the following evaluation.
l 2.0 EVALUATION l
Three areas are affected by the proposed change to the fuel enrichment limit of the reactor core. These are: (1) the reactor core, (2) the spent fuel storage racks, and (3) the dry storage racks. Each of these areas will be addressed in this evaluation.
Reactor Core The proposed revision would change the fuel enrichment limit of Technical Specification 1.3 from 3.3 to 3.5 weight percent uranium-235.
The purpose of Part A of this Technical Specification is, among other things, to provide the nominal design parameters of the fuel assemblies and reactor core.
In particular, the fuel enrichment limit is the maximum uranium-235 pennitted for fresh fuel assemblies to be loaded into the reactor core.
The cycle-specific safety analysis performed for a given fuel cycle includes the effects of the uranium-235 fuel enrichment through the number, placement, and enrichment of fresh fuel assemblies in the reactor core as well as on such other factors as exposure distribution and placement of previously burned fuel assemblies, the number and
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The effects, therefore, of the fuel enrichment on power peaking fac-tors, reactivity coefficients, and control rod worths, for example, are inherent in the cycle-specific safety analysis performed for a given fuel cycle. The results of a cycle-specific safety analysis would be used to confirm existing reactor core Limiting Conditions for Operation (LCO) of the Technical Specifications or to provide for changes in the LCO's for the new fuel cycle. The staff concludes, therefore, that it is acceptable to change the fuel enrichment limit of Technical Specification 1.3 from 3.3 to 3.5 weight percent uranium-235 since the effects of fuel enrichment are inherent in the cycle-specific safety analysis performed for a given fuel cycle and are, moreover, reflected in the reactor core's LCO's.
Spent Fuel Storaae Racks The licensee had previously performed analyses for the present Maine Yankee, spent fuel storage racks for a maximum fuel enrich-ment of 3.5 weight percent uranium-235. Both standard Maine Yankee fuel assemblies and compacted fuel assemblies were considered. The staff reviewed these anal 16, 1982 (Ref. 2)yses and issued a Safety Evaluation (SE) on June This SE concluded that the licensee's proposal to change the fuel enrichment limit of Technical Specification 1.3 to 3.3 weight percent uranium-235 was acceptable.
The staff, in that SE, also accepted the licensee's analysis of the spent fuel storage racks at the uranium-235 enrichment of 3.5 weight percent uranium-235 and concluded that the applicable criterion on the effective neutron multiplication factor (K including ap-propriateuncertainties,wasmetforbothnomib)s,pentfuelpool conditions and postulated accidents. Based on its review of the SE, the staff concludes that this previous SE supports the conclusion that the fuel enrichment limit of the spent fuel storage racks can be increased from 3.3 to 3.5 weight percent uranium-235.
Dry Storaae Racks By letter (Ref. 3) dated December 8,1986, the licensee transmitted a submittal (Ref. 4) which provided a criticality analysis of the Maine Yankee new fuel storage vault. The criticality analysis used the NITAWL-KEN 0 methodology with the 123 group XSDRN cross section library.
This calculational methodology and cross section library had previously been benchmarked against critical experiments for the spent fuel storage rack criticality analysis (see Section 2.1.2 of previous staff SE (Ref. 2)).
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Additional benchmarking of the NITAWL-KENO methodology with contin-uous energy Monte Carlo and the latest cross section data is reported in Appendix A of Reference 4.
The continuous energy Monte Carlo code used was SAM /CE with ENOF/8-V cross sections. The benchmark cases consisted of a single fuel assembly mocked-up in three-dimensional geometry and immersed in water.
The void content of the water was varied, for the different benchmark cases, from 0% to 97.5 percent voids.
In one of the benchmarks, the concrete wall surrounding the storage vault and three partial assemblies cut along their centerlines were modelled to more realistically treat radial as well as axial leakage. Although some differences are apparent in the results obtained with the two calculational methodologies due, in part, to' the KENO versus SAM /CE models and the 123 group versus the ENDF/8-F cross section data, the results indicate that the NITAWL-KENO methodology is acceptable for the calculation of K for the Maine g
Yankee dry storage racks.
The licensee performed calculations for the racks containing fresh Maine Yankee fuel assemblies with a fuel enrichment of 3.5 weight percent uranium-235 as a function of water / void.
For the fully flooded dry storage racks, the licensee calculated K to be 0.857includinganuncertaintyallowanceof0.013at'[$e95/95 probability / confidence level.
Therefore, the licensee meets the NRC criterion for the fully flooded dry storage racks that the K
including appropriate uncertainties, is less than or equal f
t8$.,95.
For the case of low-density moderation, at a water density of 0.05 grams / cubic centimeter (95% voids), the licensee calculated K to be 0.889, including an uncertainty allowance of 0.011 at I$d 95/95 probability / confidence level.
Therefore, the licensee meets the NRC criterion for the low-density (optimum moderation) case that the K including appropriate uncertainties, is less than or equal to 0.0$f, The staff concludes that the storage of fresh Maine Yankee fuel assemblies having 3.5 weight percent uranium-235 enrichment is acceptable for the dry storage racks.
3.0 CONCLUSION
S The staff concludes that the fuel enrichment limit of Technical Specifi-cation 1.3 may be changed from 3.3 to 3.5 weight percent uranium-235 since the cycle specific safety analysis for a fuel cycle inherently includes the effects of fuel enrichment which are reflected in the
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reactor core's LCO's, and since both the spent fuel storage racks and dry storage racks have been analyzed by the licensee and shown to meet applicable NRC criteria.
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4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or in a surveillance requirement. The staff has determined that the
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amendment involves 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 occupational radiation exposure. The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
5.0' CONCLUSION 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 Comission's regulations, and the issuance of the amendment will not be inimical to r
the comon defense and security or to the health and safety of the public.
Date: February 9,1987 j
Principal Contributor: D. Fieno 1
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4.0 REFERENCES
1.
Letter (MN-86-147) from J. B. Randazza (MYAPCO) to NRC (Director, NRR), dated November 25, 1986.
2.
Letter from R. A. Clark (NRC) to J. H. Garrity (MYAPCO),
dated June 16, 1982.
3.
Letter (MN-86-158) from G. D. Whittier (MYAPCO) to NRC'(Director,NRR),datedDecember8,1986.
4.
" Maine Yankee New Fuel Storage Vault Criticality Analysis,"
D. G. Napolitano and A. S. DiGiovine, YAEC-1579, Yankee Atomic Electric Company, December 10, 1986.
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