ML20076E121
| ML20076E121 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/01/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076E119 | List: |
| References | |
| NUDOCS 9108150192 | |
| Download: ML20076E121 (4) | |
Text
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S AFETY EVALVA,T10N,p,Y,,T,H,E,,0((10,E,0[, pVC LE AR, R,L ACTOR, R,E,G,Ul,A, TION RELAT1HG TO AMEt4DMENT tiOS.128 AND 107 TO FACILITY OPERAT1tiG LICEtiSE t105. OPR-70 At40 DPR-75 P,Up,LJC,5,EPVICE EL ECTRIC & GA5 COMPANY P 41 L AD E L P P,J,A,,E,L,E C T pj,C,,C p,Mf,Ap,Y DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY E,LE,CT,RJ,C,,00fPMQ SALEM I:UCLEAR GEllERATil:G STATION UNIT I405. 1 AtID 2 DOCKET li0,, 5,0,,Q2, A!!p,,50,-311 1.0 JNTRODUCTION By letter dated flovember 19, 1990, as supplemented April 1, 1991 May 20, 1991, and June 14 1991 Public Service Electric & Gas Company, Philadelphia Electric Company, DeImarva, Power and Light Company and Atlantic City Electric Company (the licensee) submitted a request for changes to the Salem tiuclear Generating Station, Unit Hos. I and 2, Technical Specifications (TS). The requested changes would revise TS 5.3.1 and 5.6.3 by removing the current maximum U-235 enrichment limit. TS 5.6.3 would be revised to allow storage of Westinghouse Standard or Vantage 5H fuel with maximum enrichment of 4.55 weight percent (w/o) U-235 provided that the reference infinite multiplication factor (k-infinity)forthefuelassembliesbelessthanorequalto1.453in unborated water at 6P' F in core geometry. TS 5.6.1 would also be revised to reflect the reactivity uncertainty associated with the use of integral fuel burnable absorber pins.
2.0 QALVATION The Salem Units 1 and 2 spent fuel pools were previously analyzed for the storage of Westinghouse 17x17 Standard fuel assemblies with enrichments up to 4.05 weight percent (w/o) U-235. The current analysis considers the storage of Westinghouse vantage 5H (V5H) fuel containing integral fuel burnable absorbers (1FBAs) with enrichments up to 4.55 w/o U-235. The fuel assembly IFBAs consist of a thin boron coating on the outside of the fuel pellet, thus making it an integral part of the fuel assembly. The primary mechanical difference between the Standard and V5H fuel is that the V5H design employs rircaloy rather than inconel spacer grids.
However, since the fuel storage criticality analyses are performed by replacing the grid volume with water volume, the results are applicable to both fuel types.
9108150192 910001 ADOCK 050 % 2 DR
l 2-t The neutron cross sections were generated with the PSCpH code, which is a depletable, two-dimensional, multigroup, transport theory code based on the Epp.1 code CpH 2.
The reactivit code, a depletable, few-group, y calculatias were performed with the PDQ7 diffusion theory code.
The analytical methods and mcdels used in the reactivity analysis have been benchmarked against c-xperimental data and industry standard codes and have been found to adequately reproduce the critical values.
The staff finds these methods and models to be acceptable.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent confidence level (95/95 probability / confidence) probability at a 95 percent that the effective multiplication factor (k-eff) of the fuel assembly array will be no greater than 0.95. Two analytical techniques are used to ensure the criticality criterion for the storage of IFBA fuel in the Salem storage raco. The first method uses reactivity equivalencing to establish the poison material loading required to meet the criticality limits.
The recond method uses the fuel assembly infinite multiplication factor (k-infinity) to establish a reference reactivity.
The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods.
A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the same k-eff when the fuel is stored in the spent fuel racks. The results show that the rack reactivity of fuel with 60 IFBA rods with an initial enrichment of 4.55 w/o U-235 is equivalent to the rack reactivity of fresh (unirradiated) fuel having an initial U-235 enrichment of 4.05 w/o V 435 and containing no IFBA rods. This equivalence relationship assures the maximum k-eff will be calculated since depletion calculations performed by the licensee have shown that the maximum reactivity of the Westinghouse fuel assemblies occurs at zero burnup for any number of IFBA rods per assembly. This method of reactivity equivalencing has been used by other licensees for fuel storage analyses and has been accepted by the staff.
The resulting k-eff for the Saiem spent fuel storage racks was less than 0.95 and included all appropriate biases and uncertainties at a 95/95 probability /
confidence invel.
This meets the NRC acceptance criterion and is, therefore, acceptable.
In order to sin.plify verification of acceptability for storage of fuel ir the spent fuel racks, a k-infinity for a fresh 4.05 w/o U-235 fuel assembly was determined. As mentioned earlier, this is equivalent to the reactivity of a 4.55 w/o U-235 fuel assembly with 60 IFBA rods. When k-infinity is used as a reference reactivity point, the need to specify an acceptable enrichment versus number of 1FBA rods correlation is eliminated. - Calculation of k-infinity for a fuel array of 4.05 w/o fuel in the Salem reactor core geometry resulted in a reference value of 1.453.
The licensee has shown that Westinghouse Standard or V5H fuel with a maximum U-235 enrichment of 4.55 w/o and a reference k-infinity of 1.453 results in a maximum k-eff of less than 0.95 when stored in the Salein spent fuel storage racks. Therefore, the only requirement needed to ensure that the fuel racks are maintained at a k-eff below 0.95 for these fuel types
.' is to verify that for each essembly the k-infinity is no greater than 1.453 at 68' F in the core geometry and the maximum enrichment is no greater than 4.55 w/o U-235.
It is possible to postulate events which could lead to an increase in storage rack reactivity. Powever, these criticality accidents for 4.05 w/o U-235 fuel have been analyzed previously and the consequences have been found to be acceptable. These conclusions lilewise apply to 4.55 w/o U-235 fuel with a sufficient r>urnber of IFCA rods to maintain an unborated reference fuel assembly k-infinity less than or equal to 1.453 at 68' F in core geometry as explained above.
Reanalysis of the spent fuel pool heat loads and radiological consequences of potential fuel handling accidents were not required to be performed as part of this submittal.
These are issues which are affected by extended fuel burnups and operationel history.
Although it is possible to achieve extended cycle burnups using assemblies with increased enrichment, actual assembly burnups will depend on core reload designs and integrated power history.
Therefore, the impact of extended burnups will be addressed as a separate issue, and any required safety analysis and FSAR updates will be performed as necessary.
Prior to this, reload cycle designs and reactor operation will ensure that the design bases of the spent fuel pool will be maintained.
Based on the above evaluation, the staff concludes that the storage racks in the Salem 1 and 2 spent fuel pools can accommodate Westinghouse Standard or V5H fuel assemblies with maximum enrichments of 4.55 w/o U-235 provided that j
the fuel with enrichment greater than 4.05 w/o U-235 contains sufficient IFBAs such that the maximum core geometry k-infinity of these assemblies is no greater than 1.453 at 68' F.
Since the analyses were performed for fuel with enrichment up to 4.55 w/o U-235, this enrichment will be included in the TS to indicate that any higher enriched fuel will require an evaluation to determine if the k-infinity criterion remains valid.
Although che Salem TS have been modified to specify acceptable reload fuel for storage, evaluations of reload core designs (using any enrichment) will of course, be performed on a cycle by cycle basis as part of the reload safety evaluation process.
Each reload design is evaluated to confirm that the cycle core detign adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.
3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENT AL_,C,0NS10ERMT10!{
pursuant to 10 CFR 51.21, 51.32 and M.35, an environmental assessment and finding of no significant impact has been prepared and published in the j
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=4-l FederalRegister(56FR33045)datedJuly 18, 1991. Accordingly, based upon the environmental assessment, the Consnission has determined that the issuance of the amendments will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
The Connission 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 Connission's i
regulations, and (3) the issuance of the amendments will not be inimical to the conanon defense and security or to the health and safety of the public, principal Contributors:
L. Kopp J. Stone Date:- August 1, 1991
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