ML20127G906

Forwards SE Concluding That Requirement for Maintaining 1 PSI drywell-torus Differential Pressure During Performance of in-plant Safety/Relief Valve Testing May Be Relaxed Provided Listed Conditions Met

ML20127G906
Person / Time
Site:	Monticello
Issue date:	12/02/1977
From:	Stello V Office of Nuclear Reactor Regulation
To:	Mayer L NORTHERN STATES POWER CO.
References
NUDOCS 9211170458
Download: ML20127G906 (3)

Text

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D, UNITED STAf tB 84

,p NUCLEAR REGULATORY commits 10N

,a T W ASHING T ON, o. C. 20966 December 2.1977

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Docket No. 50-263 Northern States Power Company ATTN: Mr. L. O. Mayer. Manager Nuclear Support Services 414 Nicollet Hall - 8th Floor Minneapolis Minnesota 55401 Gentlemen:

By letter dated August 29. 1977 you submitted a document entitled t

" Removal of Drywell-Wetwell Differential Pressure Controls" for the Monticello Nuclear Generating Plant. You stated that in an 23, 1977 meeting. the NRC staff had informed you that, as August prerequisites to removal of the differential pressure control. you must either demonstrate compliance with the criteria of the NRC's May 19,1977 letter or show that torus loads are less than allowable by code. The drywell-wetwell differential pressure control had been established in accordance with the NRC letter dated February 27 1976, and was committed to in your letter of March 1.1976.

The NRC staff still has the matter of long term removal of ap controls under review. We anticipate that the review should be completed in the near future.

In telephone conversations with the NRC staff, you have indicated your intent to conduct, on behalf of the Mark I-owners group, further in-plant safety / relief valve tests 'at Monticello in the near future. On the basis of information presented in your letter dated May 10,1976 and on the further information in your August 29, 1977.

letter regarding the in-plant safety / relief valve -testLwe have concluded that the requirement for maintaining a -1 PSI drywell-torus differential pressure during performance of in-plant safety / relief valve testing may' be relaxed provided:

1.

The differential pressure is restored during any unscheduled l

interruption of the testing which is anticipated to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

The differential pressure is restored within.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a test is completed, i

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9211170458---771202

- PDR ADOCK 05000263 P

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Northern States Power Company 2-December 2,1977 The basis for this conclusion was presented in our Safety Evaluation dated May 21, 1976, a copy of which is enclosed. Our initial review of your August 29, 1977 submittal confirms this evaluation.

Sincerely, i

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Victor Stello, Jr., Director Division of Operating Reactors Office of Nuclear Reactor Regulation

Enclosure:

Safety Evaluation dated May 21, 1976 cc w/ enclosure:

See next page I

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Northern States Power Company 3

December 2,1977 1

cc W!d Charnof f. Esquire The Environmental Conservation Library enclosure Gera Minneapolis Public Library 4

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Shaw, Pittman, Potts and 300 Nicollet Mall l

Trowbridge Minneapolis, Minnesota 55401 1800 M Street,-N. W.

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Washington, D. C.

20036 Arthur Renquist, Esquire Vice President - Law Northern States Power r.ompany l

414 Nicollet Mall Minneapolis, Minnese 65401 Mr. L. R. Eliason Plant Manager Monticello Nuclear Generating Plant i

l Northern States Power Company i

Monticello, Minnesota 55352 j-l Russell J. Hatling, Chairman Minnesota Environmental Control i

Citizens Association (MECCA) i l

Energy Task Force 144 Helbourne Avenue, S. E.-

j Minneapolis, Minnesota 55414 Mr. Kenneth Dzugan Environmental Planning Consultant l

Office of City Planner I

Grace Building 421 Wabasha Street l

St.-Paul, Minnesota 55102 i.

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Sandra S. Gardebring Executive Director

,i Minnesota Pollution Control Agency l

1935 W. County Road 82 l

Roseville, Minnesota 55113 Mr. Steve Gadler

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2120 Carter Avenue St. Paul, Minnesota -_55108 l

Anthony Z.~Roisman, Esquire Sheldon,- Harmon & Roisman 102515th Street, N. W., 5th Floor Washington, D. C.

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4 UNITED STATES

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t NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20656 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATI0'.(

SUPPORTI1?G TCIPORARY RELAXATION OF THE DRYWELL-TORUS DIFFERCITIAL PPISSURE REQUIREMENT AT MONTICELLO HUCLEAR GC ERAT 121G PLA'.T FOR TESTING PURPOSES DOC)TT NO. S0-263 INTRODUCTION By letter dated May 10, 1976, Northern States Power Company (NSP) submitted plans for the performance of an In-Plant Safety / Relief Valve Test at the Monticello Nuclear Generating Plant during the week of May 23, 1976.

To provide mesnf ngful results, such testing will require an " unbiased" initial veter level in the relief valve blowdown line and thus the temporary (approximately 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) cessation of the 1 PSI drywell-torus differential f

pressure voluntarily imposed by NSP af ter the February 26, 1976 meeting r

of the BWR Mark I owner's group and the NRC staf f.

s BACRGROLUD e.

j A January 6,1976 letter from Mr. Ivan F. Stuart of General Electric to Mr. R. S. Boyd of the USNRC described an In-Plant Safety / Relief Valve Test to be perfor=ed as part of the long-term program for evaluation of Mark I containrent systems. NSP agreed to perform the test at the Menticello Nuclear Generating Plant. Testing was originally to have been perfor:ed im March 1976. However, reanalysis of the torus support downward load /

strength ratios in February 1976' indicated that the margins of safety of the torus structure were not as large as had been previously calculated.

Thus, at the meeting of February 20, 1976 NSP, along with other Mark I licensces, agreed to establish the differential pressure between the drywell.and the torus to provide a reduction in the potential loads during a postulated loss of coolant accident and an associated restoration of the margins of safety to obtain a f actor of safety of about two.

This agreement was confirmed in a February 27, 1976 letter from Mr. Benard C. Rusche, Director, Offide of Nuclear Reactor Regulation..NRC, to_NSP.

Subsequently, USP has undertaken a program of structural modifications to enhance the-safety margins of the Monticello torus. A description of the modificatiens and the resulting loading tables are included'as.an attachtent ~to NSP's May 10, 1976 letter.

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. DISCUSSI0!! A'TD EVAI.UATIO!I We NRC staff agrees with NSP and the Mark I owner's group that the In-Plant Safety / Relief Valve Test vill provide data essential for the development of lo.ng-term solutions to the Mark I problem.

Our concern

, is mainly that there.is reasonable assurance that the torus would remain intact and function as intended should the extremely low probability loss-of-Coolart Accident (LOCA) occur during the relatively short period of time (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) that the differential pressure would be relaxed.

In their May 10, 1976 letter, NSP committed to restore the differential pressure should the test be interrupted for periods anticipated to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

NSP has submitted data, as an attachment to its May 10, 1976 letter, which supports the NSP conclusion that the modifications, which have been completed, produce a' 46% improvement in rati' of load to ultimate capacity without the differential pressure.

Improvement of over 100%

is claimed for up-loads. We have completed a preliminary review of data which has been provided and agree with NSP's statement concerning the increased safety margins with the modifications installed and the; differential ?ressure removed. Additionally, we concur that conditi6ns

• under which the test will be performed "will provide safety margins conservatively within and consistent with the margins of safety discussed in Mr. Rusche's letter of February 27, 1976" (i.e., factors of safety of about two).

The test does not involve an unreviewed safety question in that (1) the probability of occurrence er.the consequences of an accident or mal-function of equipment important to safety previously evaluated in the safety analysis report is not increased, (2) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created, and (3) the margins of safety as defined in the bases for the technical specifications are not reduced.

On5he"b' asis of the above considerations, we have concluded that elimination of the dryvell-torus pressere dif ferential= during the relatively shorp period

.of time required to perform the In-Plant Safety / Relief Valve Test is acceptable.

Date: May 21, 1976 9

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