ML20044H327
| ML20044H327 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/28/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H324 | List: |
| References | |
| NUDOCS 9306080188 | |
| Download: ML20044H327 (5) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 175 AND 178 TO FACILITY OPERATING LICENSE NOS. DPR-44 and DPR-56 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION. UNIT NOS. 2 AND 3 DOCKET NOS. 50-277 AND 50-278
1.0 INTRODUCTION
By letter dated February 5, 1993 (Reference 1), the Philadelphia Electric Company, Public Service Electric & Gas Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) submitted a request for changes to the Peach Bottom Atomic Power Station, Units 2 and 3, Technical Specifications (TS). The proposed Technical Specification (TS) changes would allow the use of the maximum k-infinity based method of demonstrating compliance with fuel storage criticality limits.
This method would replace the current limit on fuel assembly average U-235 loading.
The proposed changes would allow the storage of fuel with a calculated incore X-infinity (k.) of 51.362, in the spent fuel storage pools (SFSP) at the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.
PBAPS, Units 2 and 3 (PB-2/3) are General Electric (GE) designed boiling water reactors (BWRs).
The storage of GE 7x7 and 8x8 fuel rod lattice array assembly types has previously been reviewed and approved (Reference 2) for the spent fuel pool storage racks.
The existing TS 5.5, MFuel Storage," states that the SFSP racks are capable of storing fuel having a fuel assembly average loading not to exceed 17.3 grams of U-235 per axial centimeter of total active fuel height of the assembly.
This limit was based'on a uniform 3.50 weight percent U-235 enrichment for the original GE 7x7 fuel design.
The licensee request would revise TS 5.5.D to allow the k, method. To support this proposed change, the licensee has included a GE analysis (Reference 3) of the PBAPS SFSP racks, which addresses the conversion from a grams of U-235 loading limit to an incore k, limit, which also applies to GE 9x9 fuel.
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2.0 EVALUATION The design basis for preventing criticality in fuel storage facilities is based on Section 9.1, " Fuel Storage and Handling," of the NRC Standard Review Plan (SRP), NUREG-0800 (Reference 4).
Section 9.1.1, "New Fyel Storage," effectively requires, by reference to ANSI Standards ANS 57.1 (Reference 5) and ANS 57.3 (Reference 6), that there is a 95 percent probability at a 95 percent confidence level (95/95 probability /
confidence) that the effective multiplication factor (k,,,), including uncertainties, will be no greater than 0.95 under unborated moderator conditions and no greater than 0.98 under optimum moderation.
Section 9.1.2, " Spent Fuel Storage," requires, by reference to ANS 57.2 (Reference 7), at a 95/95 probability / confidence level, that k for fuel stored under normal conditions or accident conditions (such as,,a, dropped fuel assembly) shall be less than or equal to 0.95.
Credit may be taken for integral burnable poisons in the fuel and for fuel burnup effects.
General Design Criterion (GDC) 62 (Reference 8) also states that:
" Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations".
2.1 Criticality Analysis Methods An analysis of.he criticality aspects of the storage of PBAPS fuel assemblies having a fuel enrichment up to 4.5 weight percent U-235 was performed by the licensee's current fuel vendor (GE).
The analysis methodology and results were described in the GE report that was furnished as Attachment 3 to Reference 1.
The method and the cross-section library used consist of the GE MERIT computer code, using the ENDF/B-IV cross section set, which is stated to have been verified against extensive critical experiments. The MERIT program is a three-dimensional Monte Carlo neutron tracking code, that calculates the system effective neutron multiplication factor (k,,,).
The MERIT code uses a 190 group cross-section library with the Haywood scattering kernel for water.
The GE MERIT methodology has been extensively benchmarked against critical experiment measurement data to determine the calculational uncertainty and bias for specific applications.
The benchmark measurements included six standard sets of measured critical experiments which verified the application of the MERIT code and ENDF/B library.
The results of the comparisons yielded a MERIT calculational bias of 0.0054 (10.0022) delta-k for water moderated uranium lattices.
4.
4 e 2.2 f_qe1 Storaae Rack Analysis The SFSP high-density storage racks consist of 15 rack modules with storage cells on a nominal center to center distance of 6.28 inches in both directions as described in the PBAPS Updated Final Safety Analysis Report.
The criticality of fuel assemblies in the storage racks is prevented by limiting the fuel reactivity, by the Boraflex neutron absorbers between the storage cells, and by maintaining a minimum separation between assemblies. The NRC acceptance criteria that fuel assembly storage must meet is that the k shall be no greater than 0.95 when the racks are fully loaded and flooded with pure, unborated water.
The k shall include all biases and uncertainties at a 95/95 probability /confidenc,e,, level.
The GE analyses showed that a fully loaded SFSP would meet the NRC acceptance criterion of k,#' hat this rack configuration can safely accommodate up to 4.5 less than 0.95 under flooded conditions. A conservative analysis shows t weight percent U-235 fuel with a maximum 95/95 storage rack k,,,K, 5; 1.362.
of 0.918, resulting in a fuel storage compliance limit of maximum incore This meets the staff acceptance criterion for k n
also satisfies GDC 62, and is therefore acceptaLVe.o greater than 0.95 and 2.3 Accident Analysis Certain postulated events which could lead to a storage rack reactivity increase were evaluated in the high-density rack analysis of Reference 2.
A dropped fuel assembly on top of the rack will be sufficiently separated from the active fuel height of the assemblies in the rack such that there will be no storage rack reactivity increase.
Conditions which would result in an increase in reactimty such as dropping or misloading a fuel assembly outside or adjacent to the r uk were also evaluated. The evaluation showed that an assembly dropped or misloaded in a maximum reactivity configuration meets the staff acceptance criterion of k,,, no greater than 0.95 under any condition.
2.4 Summary Based on the above review, the staff concludes that fu'el assemblies having a maximum incore k, of 1.362 may be stored in the fuel storage racks and that TS 5.5, " Fuel Storage," may be revised as proposed by the licensee in the February 5, 1993 application. Our conclusion is based on the following:
1.
The criticality analyses involved in this change have been performed with a methodology which has been extensively benchmarked by the fuel vendor against industry standard critical experiments.
2.
Appropriate uncertainties have been accounted for at the 95/95 probability / confidence level.
3.
Abnormal events and accidents that have been previously considered are not affected by the change in the reactivity basis.
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The effective neutron multiplication factor, including uncertainties, meets our acceptance criteria for all postulated conditions.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified.of the proposed issuance of the amendments.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area'as defined in 10 CFR Part 20. The NRC 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 occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (58 FR 12766). Accordingly, the amendments meet 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 the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, i
that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
E. Kendrick
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Date: thy 28,1993 f
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REFERENCES 1.
Letter from Mr. George J. Beck, Philadelphia Electric Company, to Document Control Desk, U.S. NRC, dated February 5,1993, " Peach Bottom Atomic Power Station, Units 2 and 3, Technical Specification Change Request."
2.
" Safety Evaluation by Office of Nuclear Reactor Regulation Supporting Amendment Nos. 116 and 120 to facility Operating License Nos. DPR-44 and DPR-56," for Peach Bottom Atomic Power Station, Units 2 and 3, Docket Nos.
50-277 and 50-278, dated February 18, 1986.
3.
GENE-512-92073, " Peach Botcom Atomic Power Station Spent Fuel Storage K-infinity Conversion Analyses," General Electric Nuclear Energy, November 1992.
4.
U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 9.1,
" Fuel Storage and Handling," NUREG-0800 (Revision 2), July 1981.
5.
ANS 57.1/ ANSI-N208, " Design Requirements for Light-Water Reactor Fuel Handling Systems."
6.
ANS 57.3, " Design Requirements for New LWR Fuel Storage Facilities."
7.
ANSI /ANS-57.2-1983, " Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants."
8.
U.S. Code of Federal Reaulations, Title 10, Chapter I, Part 50, Appendix A, General Design Criterion 62, " Prevention of Criticality in Fuel Storage and Handling."
9.
U.S. Nuclear Regulatory Commission, Regulatory Guide 5.14, " Validation of Calculational Methods for Nuclear Criticality Safety."
- 10. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.13, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations."
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