ML20045G267
| ML20045G267 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/28/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045G261 | List: |
| References | |
| NUDOCS 9307130115 | |
| Download: ML20045G267 (6) | |
Text
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UNITED STATES 1 I NUCLEAR REGULATORY COMMISSION e
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.166 T0.
FACILITY OPERATING LICENSE N0. DPR-51 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NO. 1 DOCKET NO. 50-313
1.0 INTRODUCTION
By letter dated June 27, 1991 (Reference 1), Entergy Operations, Inc,-(E01),
the licensee, submitted proposed Technical Specification (TS) changes which would permit fresh, unirradiated fuel having an initial enrichment of up to 4.1 weight percent U-235 to be stored in the fresh fuel storage racks (FFSRs) for use in the core of Arkansas Nuclear One, Unit No. 1 (ANO-1).
ANO-1 is a Babcock & Wilcox (B&W) designed pressurized water reactor (PWR),
and B&W 15 x 15 fuel rod lattice array fresh fuel of this enrichment has previously been reviewed and approved (Reference 2) for storage in the spent fuel pool storage racks.
However, the existing AN0-1 TSs state that the FFSRs are capable of storing fuel having an enrichment no greater than 3.5 weight percent U-235. The licensee's request would revise TS 5.3.1.6, " Reactor Core," and TS 5.4.1.1, "New Fuel Storage", to increase the maximum initial U-235 enrichment of future reload fuel being cycled through the facility from 3.5 to 4.1 weight percent.
The requested changes.to TS 5.4.1.1 also include the addition of Figure 5.4-1, "AND FFSR. Loading Pattern," to indicate the locations in the FFSR that will be prohibited from use. To support these proposed changes, the licensee has included a criticality analysis (Reference 3) of the AN0-1 FFSR which addresses the storage of 4.1 weight percent U-235 fuel assemblies, using the same methodology (Reference 4) which-was previously accepted for the Unit.No. 2 (AN0-2) FFSR criticality analysis (Reference 5),
2.0. EVALUATION The design basis for preventing criticality in new fuel storage is based on.
the NRC Standard Review Plan (SRP), NUREG-800 (Reference 6)..Section 9.1.1, "New Fuel Storage," effectively requires (by reference to the. ANS 57.1 (Reference 7} and ANS 57.3 { Reference 8} standards) that there is a 95 percent probability at a 95 percent confidence level (95/95 probability / confidence) i that the effective multiplication factor (k-effective), including uncertainties, will be no greater than 0.95 under unborated moderator conditions and no greater than 0.98 under optimum moderation.
General Design Criterion (GDC) 62 (Reference 9) also states that:
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. 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 The analysis of the criticality aspects of the storage of ANO-1 fresh fuel assemblies having a fuel enrichment of 4.1 weight percent U-235 was performed by the licensee. The analysis methods used consist of the AMPX/NITAWL/ KEN 0 computer codes, which are part of the Oak Ridge National Laboratory (ORNL)
Standard Computer Analysis for Licensing Evaluations (SCALE) code package (Reference 10).
The AMPX code uses the 123 group SCALE cross-section library with the NITAWL routine to derive weighted cross-sections for U-238 in the resonance region with the Nordheim resonance integral treatment.
The NITAWL output is used by the KEN 0 program, a three-dimensional Monte Carlo neutron tracking code, that calculates the system effective neutron multiplication factor (k,,,).
The AMPX/ KEN 0 methodology has been extensively benchmarked by the nuclear industry, including the current fuel vendor (B&W Fuel Company).
However, the licensee has also performed additional critical experiment benchmarks with their own specific methodology to determine the calculational uncertainty and bias f.or their specific applications.
The benchmark measurements included a standard set of B&W critical experiments (Reference 11) and a set of 70 shipping cask critical experiments at ORNL (Reference 12) which verified the application of the SCALE library.
The results of the licensee comparisons from Reference 4 yielded a KEN 0 one-sided upper tolerance limit of 0.021 with a probability of 95% that at least 95% of the calculated KEN 0 results will be within this limit (95/95 probability / confidence level).
Another industry standard code, CASH 0, is used to determine the reactivity ef fects of variations in fuel pellet density and U-235 enrichment.
CASMO is a multigroup two-dimensional transport theory code which has been validated by the licensee for calculations of PWR fuel lattice reactivities (Reference 13).
2.2 New Fuel Storaae Rack Analysis The FFSR consists of a nine-by-eight array of storage cells on a nominal center to center distance of 21 inches in both directions as described in the ANO-1 UFSAR.
The criticality of fuel assemblies in the ANO-1 FFSRs is prevented by limiting the U-235 enrichment in the fuel rods to 4.1 weight percent and by maintaining a minimum separation of 21 inches between assemblies.
Since unirradiated fuel contains no radioactive fission products, it requires no cooling or shielding and is normally stored in a dry condition.
However, the NRC acceptance criteria that fuel assembly storage must meet are that the k, shall be no greater than 0.95 when the racks are fully loaded and floodeIJ,with pure, unborated water and that the k,,,
immersed with low-shall be no greater than 0.98 under " optimum" moderation when the racks are density hydrogenous material due to such causes as mist, fog, or fire-fighting
' foam.
The k shall include all biases and uncertainties at a 95/95 probabNity/ confidence level.
The licensee has performed calculations for the FFSRs at various moderator densities in order to obtain the optimum moderator density which results in the maximum reactivity.
Since AN0-1 fresh fuel is stored in cavities whose internal dimensions are larger than the outside fuel assembly dimensions, there is an uncertainty in the lateral placement of any one assembly in its cavity.
However, the licensee has previously shown that eccentric placement of assemblies reduces the multiplication factor in optimum moderation analyses.
The assemblies, therefore, were modeled as centered in the cells.
The licensee analyses showed that a fully loaded FFSR (72 assemblies) would meet the NRC acceptance criterion of k less than 0.95 under flooded conditions;however,inordertomeetINeNRCacceptancecriterionofk,,,
less than 0.98 for conditions of optimum moderation, it is necessary to preclude (by physical blockage) the placement of fresh fuel assemblies in 10 interior storage cell locations as shown in the proposed Figure 5.4-1.
A conservative analysis shows that this partially loaded rack configuration can safely accommodate 4.1 weight percent U-235 fuel with a 95/95 probability / con-fidence level for k of 0.96957 under conditions of optimum moderation.
Thismeetsthestaff,acceptancecriterionfork also satisfies GDC 62, and is, therefore, accepla# no greater than 0.98 and bl e.
2.3 A_.cident Analysis Certain postulated events which could lead to a storage rack reactivity increase were evaluated.
Asymmetric positioning of fuel assemblies in the cells has been shown to yield equal or conservative results compared to symmetrically positioned fuel assemblies.
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 reactivity such as dropping or misloading a fuel assembly outside or adjacent to the rack were evaluated for the normally dry rack. This is acceptable by application of the double-contingency principle of ANSI N16.1-1975, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," which states that the evaluation is not required to assume two unlikely, independent concurrent events to provide for protection-against a criticality accident.
The evaluation showed that an assembly dropped or misloaded in a maximum reactivity configuration beside the normally dry FFSR results in a 95/95 k 0.546718, well below the staff acceptance criterion of k,,, no greater th,a,n, of 0.95.
In our review and evaluation of the proposed changes, the following considerations apply:
1.
The criticality analyses involved in this change were performed with an industry standard methodology which has been additionally benchmarked by the licensee.
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2.
Appropriate uncertainties were accounted for at the 95/95 probability / confidence level.
3.
Abnormal events and accidents were considered.
4.
The effective neutron multiplication factor, including uncertainties, met our acceptance criteria for all postulated conditions.
Based on our review, we conclude that up to 62 fresh, unirradiated fuel assemblies having a maximum initial enrichment of 4.1 weight percent U-235 may be stored in the ANO-1 fresh fuel storage racks and that TS 5.3.1.6, " Reactor Core," and TS 5.4.1.1, "New Fuel Storage," may be revised accordingly, including the addition of Figure 5.4-1 to indicate the physically blocked storage locations.
3.0 STATE C_0jjSVLTATION In accordance with the Commission's regulations, the Arkansas
- State official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
The NRC staff has determined that the 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 Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (56 FR 37580). Accordingly, the 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 need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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. REFERENCES 1.
Letter from N. Carns (Entergy Operations) to Document Control Desk (USNRC), dated June 27, 1991, " Proposed Change to the Technical Specification for Increased Fresh Fuel Enrichment."
2.
Letter from J. Stolz and R. Clark (NRC) to J. Griffin, (AP&L) dated April 15, 1983, issuing Amendment No. 76 to Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit No. I and Amendment No. 43 to facility Operating License NPF-6 for Arkansas Nuclear One, Unit No 2.
3.
" Criticality Analysis of ANO-1 Fresh fuel Rack," by M. R. Eastburn, Entergy Operations, Inc., December 1990.
4.
Letter from J. Enos (AP&L) to G. Knighton (NRC), dated January 29, 1986, submitting "ANO-2 Fresh fuel Pit criticality Analysis, Storage of 4.1 Weight Percent U-235 Assemblies," by M. R. Eastburn, Middle South Services, Inc., October 17, 1984.
5.
Letter from R. Lee (NRC) to J. Griffin (AP&L), dated April 16, 1986,
" Issuance of Amendment No. 71 to Facility Operating License NPF Arkansas Nuclear One, Unit No. 2."
6.
USNRC Standard Review Plan, Section 9.1.1, "New fuel Storage," NUREG-0800 (Rev. 2), July 1981.
7.
ANS 57.1/ ANSI-N208, " Design Requirements for Light-Water Reactor Fuel Handling Systems."
8.
ANS 57.3, " Design Requirements for New LWR Fuel Storage Facilities."
9.
Code of Federal Regulations, Title 10 Part 50, Appendix A, General Design Criterion 62, " Prevention of Criticality in Fuel Storage and Handling."
10.
R. M. Westfall, et al., " SCALE-2: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation," NUREG/CR-0200, Oak Ridge National Laboratory.
11.
N. M. Baldwin et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor fuel," BAW-1487-7, Babcock and Wilcox, July 1979, 12.
R. M. Westfall, J. R. Knight, " SCALE System Cross Section Validation with Shipping Cask Critical Experiments," Transactions of the American Nuclear Society, Vol. 33, 1979, p. 368.
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Letter from R. A. Clark and J. F. Stolz (NRC) to W. Cavanaugh (APL),
dated August 11, 1982, transmitting " Evaluation of Report MSS-NAl-P,
' Qualification of Reactor Physics Methods for Application to Pressurized Water Reactors of the Middle South Utilities System.'"
Principal Contributor:
E. Xendrick Date:
June 28, 1993 1
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