ML20234E053

Safety Evaluation Accepting Design Changes to Be Implemented During Unit 1 Cycle 7 Fuel Reload.Tech Spec Change Unnecessary

ML20234E053
Person / Time
Site:	Beaver Valley
Issue date:	12/08/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20234E036	List: ML20234E033 ML20234E053
References
NUDOCS 8801070403
Download: ML20234E053 (2)

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SAFETY EVALUATION BY THE OFFICE Of NUCLEAR REACTOR P.EGULATION BEAVER VA,L,LE_Y UNIT 1 CYCLE 7 RELO,AD 1.0 1hTROD_UCTION F4y letter dated November 12, 1987 Duquesue Light Company, the licensee for the-Beaver Valley Unit I nuclear power plant, submitted a Westinghouse document entitled "Feload Safety Evaluation, Beaver Valley Nuclear Plant, Unit 1, Cycle 7 " dated October 1987. The staff has reviewed that submittal and documented its review results as follows. '

?.0 DISCUSSION AND EVALUATION The licensee's submittal does not include any technical specification changes;-

however, the submittal describes the following design differences incorporated into the Cycle 7 reload fuel: (1) the fuel pellet stack holddown springs exert a4 (3)5Integrated axial force instead Fuel of aAbsorber Burnable 69; (2) the fuel pellets have chamfered edges; Annular Burnable Absorbers (VABA); (4)(IFBA) rods are used in lieu of Vet IFBA fuel rods contain a lower Feliut backfill pressure compared to non-IFBA fuel rods to offset the ef fects of the helium gas release from the IFBA coating during irradiation; (5) the reload fuel has 6-inch axial blankets of natural uranium pellets that are of the sane design as the enriched pellets except that these pellets are not dished; and (6) reduction of reactor flow due to a 2% increase in the bypass flow.

The new fuel constitutes region 9 of the Cycle 7 reload. The report discusses reanalyses of those incidents potentially affected by the new reload. The rearalyses were performeJ using the NRC-approved methodology described in WCAP-9272-A, 1985.

• Westinghouse Reload Safety Evaluation Methodology," issued July The Cycle 7 core reload is of a low-leakage configuration with 77 new assemblies in region 9. The region 9 fuel changes listed above were evaluated using the same rnetheds as in the Cycle 6 reload. Clad flattening is not predicted to occur and the fuel rod internal pressure design basis is satisfied for all fuel regicns. The F x P ECCS limit is satisfied and the flux difference delta I -

band is within hi as required by the technical specifications. The moderator temperature coefficient is kept negative at hot zero power by administrative controls which require partial control rod bank D insertion. The limiting fuel type assurnption for the FSAP large-break LOCA continues to apply and the IFBA fuel rod is bounded by the FSAP analyses. The Cycle 7 control rod worths exceed the required minimum shutdown margin. Drooped rod incidents were 8801070403 871229 DR ADOCK 050 g 4

evalueted using ar. approved method (WCAP-10298-A, June 1983) and the design bases are net. Likewise, the peaking factors for a st(an, line break are withir the previous eralyses limits. There are no significar,t changes in the therr.al o,aigins fer Cycle '. The core bypass flew will increase from 4.5% to 6.ET of total teacter flow due to thimble plug recioval. This ctange which results in core flow reduction has beer, evaluated and found to te acceptable, i.e., the CNB ratio is within the design limits. Finally the licensee established that in the overpower transient,the fuel center line ten.perature limit of 47CC' F is well within the limits of the Cycle 7 core,

, 3.0. CONCLUSION We conclude that the Cycle 7 design does not cause the previously acceptable limits for any incident to be exceeded, and is thus acceptable.

Pri n c i pa_1_ ,Co n,t,r,1,bu, tor Lambros Lois

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Dated: December 8,1987 i

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