ML20049H646

Discusses Basic Parameters of Reload Design for Cycle 3 & Safety & Operational Aspects of Westinghouse Design of Reload.Fuel Assemblies Will Be Replaced When Unit Returns to Power for Cycle 3 on 820523

ML20049H646
Person / Time
Site:	Beaver Valley
Issue date:	02/23/1982
From:	Carey J DUQUESNE LIGHT CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
TAC-48018, NUDOCS 8203030407
Download: ML20049H646 (4)

Text

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• • February 23,

• Director of Nuclear Reactor Regulation 3 United States Nuclear Regulatory Commission Attn: Mr. Steven A. Varga, Chief 2 '~.W ..

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Operating Reactors Branch No. 1 C; MSR2 Jggg 'g Division of Licensing A Washington, D.C. 20555 UtInt;;ym D g Ugd

Reference:

Beaver Valley Power Station, Unit No. I s ,,,

Docket No. 50-334, License No. DPR-66 @

Cycle 3 Reload Safety Evaluatics Gentlemen:

Beaver Valley Power Station, Unit Nom 1 completed its second cycle of operation on December 25, 1981, having achieved a core average burnup of 9640.2 MWD /MTU. The plant is presently shutdown for refueling and modifications. We expect to return to power operation for Cycle 3 on May 23, 1982. This letter is to advise you of the basic parameters of the reload design for Cycle 3 and the Duquesne Light Company review of the safety and operational aspects of the design of the reload by Westinghouse Electric Corporation.

The Beaver Valley Unit No. 1 - Cycle 3 reload core was designed to perform under current nuclear design limits, Technical Specifications, related bases and current Limited Safety Systems Settings applicable to this unit. The reload fuel consists of a total of fifty-two (52) standard design, Westinghouse 17 x 17 fuel assemblies, with the exception of minor grid modifications to minimize potential grid to grid interaction during fuel handling (identified as Region 5 fuel).

For Cycle 3 operation, Beaver Valley Unit No. I will replace (1)

Region 1 and (52) Region 2 fuel assemblies with (52) Region 5 assemblies and (1) Region 1 assembly discharged at the end of Cycle 1. As in Cycle 2, this cycle will contain (2) Region 4 demonstration assemblies. The design and operational characteristics of the (2) optimized demonstration assemblies are described in the Westinghouse report " Optimized Fuel Assembly Demonstration Program" (WCAP-9286), July,1978.

Duquesne Light Company has performed a detailed review of the Westinghouse Reload Safety Evaluation Report (RSER) for Beaver Valley _.

Power Station, Unit No. 1 - Cycle 3, including all postulated accidenta in the FSAR. The RSER included a review of the core characteristics to determine those parameters af fecting the postulated accident analyses reported in the Beaver Valley FSAR. The incidents whose consequences U, h/

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Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Cycle 3 Reload Safety Evaluation Page 2 could potentially be affected by the reload core characteristics were reanalyzed. Duquesne Light Company verified that the reanalyses were performed in accordance with the Westinghouse reload safety methodology as outlined in the March, 1978, Westinghouse topical report entitled,

" Westinghouse Reload Safety Evaluation Methodology" (WCAP-9272).

Under separate application for amendment to the Beaver Valley Technical Specifications we have requested that the current FAH Technical Specification limit of 1.5355 be replaced with a 1.55 limit, having rod bow accounted for by the burnup dependent expression for rod bow penalty [1-RBP(BU)) to conform to the standard method for accounting for rod bow penalty. The 0.2 power multiplier will be replaced by 0.3 in the FAH and F limits. The core limits have been evaluated for this change and it has beenXketermined that Figures 2.1-1 and 2.1-2 require revision. The proposed axial of fset DNB envelopes result in a more limiting f(AI) function; therefore, Table 2.2-1 items (1) and (iii) will be revised accordingly. A Technical Specification change has been submitted to extend to Cycle 3 and future cycles the current Technical Specification related to deviations of t16 steps between rod demand position indicators and analog rod position indicators; also, the allowance for the individual rod position indication system accuracy will be increased to !16 steps. The analysis performed to support the !32 step total rod misalignment change has shown that the peaking factor increases resulting from misalignments of 32 steps are small at significant power levels and are within the available margin and conservatisms employed in the RSER. At very low power levels, FAH and F limits may be exceeded; however, the core limits are vessel extt boiling limithdandarenotafunctionofpeakingfactors. Consequently, exceeding these peaking factor limits at low power levels does not pose a safety concern.

Plants are routinely operated with rods essentially fully withdrawn; this mitigates large upward rod misalignments which analysis has found to be much more limiting than downward rod misalignments. This, combined with the fact that actual rod misalignments are very rare in Westinghouse plants, makes the occurrence of large misalignments very unlikely. The above conclusions are valid for N-1 loop operation as well as 3 loop operation since the peaking factor characteristics of the 3 loop and N-1 loop cores are very similar.

These Technical Specification changes have been factored into the Reload Safety Evaluation, but the validity of this Reload Safety Evaluation is not dependent upon these proposed Technical Specification changes.

The Beaver Valley Onsite Safety Committee (OSC) and the Duquesne Light Of fsite Review Committee (CRC) have concluded that no unreviewed safety questions, as defined by 10 CFR 50.59, are involved with this reload.

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Beaver Valley Power Station, Unit No.1 Docket No. 50-334, License No. DPR-66 Cycle 3 Reload Safety Evaluation Page 3 The reload core design will be verified by performing the normal standard startup physics tests for Westinghouse PWR reload cycles.

These tests will include:

1. Control rod drive tests and rod droptime mensurements;

2. Critical boron concentration measurements;

3. Control rod bank worth measurement;

4. Moderator temperature coef ficient measurement;

5. Power coef ficient measurement;

6. Startup power distribution measurements using the incore flux mapping system.

Results of these tests will be submitted in accordance with Section 6.9.1.3 of the Technical Specifications.

Very truly yours, a

J. J. Carey Vice President, Nuclear

Beaver Valley Power Station, Unit No.1 Docket No. 50-334, License No. DPR-66 Cycle 3 - Reload Safety Evaluation l Page 4 cc: Mr. D. A. Beckman, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, Pennsylvania 15077 U. S. Nuclear Regulatory Commission c/o Document Management Branch Washington, D.C. 20555 1

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