B11561, Forwards Reload Safety Analysis in Support of Cycle 7 Reload.Core Reload Will Not Adversely Affect Plant Safety

Forwards Reload Safety Analysis in Support of Cycle 7 Reload.Core Reload Will Not Adversely Affect Plant Safety

ML20126J122
Person / Time
Site:	Millstone
Issue date:	06/05/1985
From:	Opeka J, Sears C NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20126J126	List: B11561, Forwards Reload Safety Analysis in Support of Cycle 7 Reload.Core Reload Will Not Adversely Affect Plant Safety ML20126J181
References
B11561, TAC-56814, NUDOCS 8506100513
Download: ML20126J122 (4)

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==es rmoma co*" HARTFORD, CONNECTICUT 06141-0270 L L J [",', [",[,"[.,,, (203) 665-5000 June 5,1985 Docket No. 50-336 Bil561 Director of Nuclear Reactor Regulation Attn: Mr. James R. Miller, Chief Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) W. G. Counsil letter to R. Reid, dated March 6,1980.

(2) R. A. Clark letter to W. G. Counsit, dated June 22,1981.

(3) R. A. Clark letter to W. G. Counsil, dated January 12,1982.

(4) R. A. Clark letter to W. G. Counsil, dated February 18,1982.

(5) W. G. Counsil letter to R. A. Clark, dated November 17, i 1983.

(6) W. G. Counsil letter to 3. R. Miller, dated February 6,1985.

(7) W. G. Counsit letter to R. A. Clark, dated November 17, 1981.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Cycle 7 Refueling - Reload Safety Analysis The attached Reload Safety Analysis (RSA) is submitted in support of the Millstone Unit No. 2, Cycle 7 reload. The RSA presents an evaluation for the Cycle 7 reload which demonstrates continued conformance to the design and safety limits of the plant, in Reference (1) Northeast Nuclear Energy Company (NNECO) provided the Staff with the Basic Safety Report (BSR). References (2), (3) and (4) document the Staff's acceptance of this report. In Reference (5) NNECO presented the Staff with the Millstone Unit No. 2, Cyc!c 6, Reload Safety Analysis. The BSR, as supplemented by Reference (5), provided the basis to which the Cycle 7 reload was evaluated.

In Reference (6) NNECO submitted proposed revisions to the Millstone Unit No.

2 Technical Specifications and the preliminary Reload Safety Analysis in support of the Millstone Unit No. 2, Cycle 7 reload. Reference (6) was submitted to provide a preliminary description of the expected characteristics of the Cycle 7 B506100513 850605 PDR ADOCK 05000336 ,

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r reload and to provide the basis for all anticipated changes to.the Technical Specifications. There are no Technical Specification changes, other than those presented in Reference (6), necessary for Cycle 7 operation.

The attached RSA presents the final evaluation for Millstone Unit 2, Cycle 7, which demonstrates that the core reload will not adversely affect the safety of the plant. The attached Cycle 7 RSA provides the results of reviews of those incidents analyzed and reported in References (1) and (5), which could potentially be affected by the fuel reload. Cycle specific parameters were examined in the areas of core kinetics, shutdown margin, Control Element Assembly (CEA) worths, and core peaking factors.

Details concerning anticipated Cycle 7 hardware changes, to be made to the lead Control Element Assembly (CEA), bank, were provided in Reference (6). The part-strength control rods in the lead CEA bank have been replaced with full-strength control rods, making the lead bank CEAs identical in composition to the

. remainder of the CEA banks. Parameters which are input to the safety analysis and which would be affected by the proposed hardware change were analyzed.

These parameters are radial peaking factor, dropped rod peaking factor, ejected rod worth and peaking factor, and shutdown margin. The only input parameter to the safety analysis which exceeded the BSR (Reference (1)) and the Cycle 6 reload safety analysis (Reference (5)) was the Hot Full Power (HFP) ejected rod -

worth.

The ejected rod accident at HFP was reanalyzed for Cycle 7 operation. The parameters used in the analysis are given in Reference (7), with the exception of the reactor coolant flow and the ejected rod worth. The reactor coolant flow used is given in the Cycle 6 reload safety analysis (Reference 5). As provided in Reference (5), the minimum reactor coolant flow rate of 350,000 gpm was established to correspond to a plugging level of 2500 steam generator tubes. The total plugging and sleeving of steam generator tubes, to date, at Millstone Unit 2 will not reduce the reactor coolant flow rate below the established minimum.

The higher worth rods, thus higher ejected worth, raised the reactivity insertion of the ejected rod accident, thus increasing the consequences of the event. The energy deposition increased from 171 calories per gram to approximately 185 calories per gram, however, the limiting criterion given in Reference (1) for this accident is not exceeded.

NNECO has determined that the increase in consequences of the HFP ejected rod accident constitutes an unreviewed safety question pursuant to 10CFR50.59(a)(2)(i). However, the limiting criterion given in Reference (1) for this accident is not exceeded; that is, the average enthalpy of the hottest fuel pellet does not exceed the damage threshold of 200 calories per gram. The actual increase in consequences is acceptable, for Cycle 7 operation, from a safety standpoint and no changes to the Technical Specifications are necessary as a result of the hardware change to the lead CEA bank.

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In summary, ' the reanalysis of the HFP ejected rod accident have been-determined to result in increased consequences for Cycle 7 operation. As such, pursuant to ' 10CFR50.59(a)(2)(i),. an unreviewed safety question exists for Millstone: Unit No. 2, Cycle 7 operation. ~ The actual' increase in consequences, however, does not exceed Technical Specification bases or FSAR criteria for

- Cycle 7. operation and are acceptable from a safety standpoint. Thus, the attached analyses supports the changes proposed in . Reference (6) and demonstrates continued conformance to the design and safety limits of the plant.

i NNECO reviewed the proposed Reference (6) changes pursuant to 10CFR50.91(a)

~and . determined ~ that the changes do not involve a significant hazards l determination. The basis for.this determination was detailed in the Reference l

'(6) submittal. Additionally, NNECO has reviewed the analyses supporting Cycle 7 operation, pursuant to the requirements of 10CFR50.91(a) and has determined

- that they do not involve a significant hazards determination. This conclusion l has been reached because the criteria delineated in 10 CFR 50.92 have not been compromised. ' That is, none of the proposed changes or analyses of Reference (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident previously evaluated; or involve a significant reduction in a margin of safety. This proposed amendment falls within -the envelope of example (vi)

- provided in 48 FR 14864 of amendments likely to involve no significant hazards L consideration. While there is an increase in consequences, all results are within acceptance criteria and w..nin the design and safety limits of the plant.

The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the

above proposed changes c'd concurred with the above determinations.

i In accordance with th'e requirements of 10CFR50.91(b), a copy of this document l 1s being provided to the State of Connecticut.

[ We remain available to assist you further in this matter such that an expeditious l review of the Reference (6) license amendment request ca'n be completed prior to startup from the upcoming refueling outage, currently estimated for June 19, 1985.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY i.

? 3. F. Opeka '

Senior Vice President L

By: C. F. Sears Vice President L

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t cc: M.. Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection State Office Building

' Hartford, Connecticut 06106 STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before me C. F. Sears, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensee herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.

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