ML20128A990

Forwards Suppl Rept Re NSP Application for OL for Proposed Plant

ML20128A990
Person / Time
Site:	Monticello
Issue date:	01/06/1970
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Hendrie J Advisory Committee on Reactor Safeguards
References
NUDOCS 9212030528
Download: ML20128A990 (12)

Text

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0FilCIALllSE 0llli JAN 6 SM Docket No. 50-263 it l L

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Dr. Joseph M. llendria Chairman, Advisory Committes

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on Reactor Safeguards U. S. Atumic Energy Cosenission Washington, D. C.

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Dear Dr. Hendriet Ei.hteen copit's of a suppleruentary report prepared by the Division t

of 11 actor Licensing are enclosed for the review of the Corumittee.

The report relates to the Northern States Power Company's applica-tit,r. for an operating license for its proposed Monticello i4uclear Gencrating Plant linit 1.

.a p Sincerely, cussani Rd M petw A, mente.

Peter A. Morris, Director Division of Reactor Licensing Lncionure :

ACRS Report (16) i Distribution:

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Docket L. 50-263 January 5, 1970' Supplemental Report to ACRS_

MONTICELLO NUCLEAR GENERATING PLANT 1

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1.0 INTRODUCTION

AND SU:2%RY In our report to.he ACRS dated November 24, 1H P._on the Monticello f acility, we reported the results of our safety evaluation.

In that-report there vere a number of items for which the results of our evaluation _ vere incotplete, primarily because we did not have sufficient confirmatory information at the time the report was written. We indicated that we veuld report to the Committee the follow-up action'on these matters. This supplemental report discusses the status of these matters:

0) Non-isolable break in the HPCI steam supply line (Section 6.1.1).

(2) Limitatien on liPCI operation (Section 4.7).

(3). Acceptability of jet pump castings (Section 4.3).

(4)

Vibratien menitoring of reactor internals (Section 4.3).:

(5)

Seismic design of structures (Section.5.1.2).

(6)

Instrutentation and control items:

ECCS low pressure auto-relief interlock (Section 9.2.3).

a.

b.

Femote testing capability of pressure switches on bellows of relief / safety valves (Section 9.2.3.3).

Single f ailure criterion (Section 9.2.5).

c.

d.

Acceatatility _of stan:iby gas treatment system initiating circuitrv (Section 9.2.4).

Seismic testing of Class -I instrumentation' (Section 9.2.7)..

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1.0 INTRODUCTION

AND SU! NARY In our report to the ACRS, dated November 24,-1969, on the Monticello facility, we reported the results of our safety evaluation.

In that report there were a number of items for which the results of our evaluation were incomplete, primarily because we did not have' sufficient confirmatory fnformation at the time the report was Vritten.

L'e indicated that 'we would report to the Committee the follow-up action-on these matters. This supplemental report discusses the status of these matters:

l (1) Non-isolable break in the llPCI steam supply line (Section '6.1.1).

(2) Limitatien on !!PCI operation (Section 4.7).

(3) Acceptability of jet pump castings ~ (Section 4.3).

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Vibration monitoring of reactor internals (Section 4.3).

(5) Seismic design of structures (Se' tion 5.1.2).

e (6) Instrumentation and control itemst ECCS low pressure auto-relief interlock (Section 9.2.3).

a.

b.

Pemote testing capability of pressure switchu on bellows of relief / safety valves (Section 9.2.3.3).

Single failure criterion (Section 9.2.5).

c.

d.

Accc-tability of standby gas treatment system initiating circuitry (Section 9.2.4).

4 Seis;aic testing of Class I instrumentation (Section 9.2.7).

c.

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OFFECHAL USE ONL J Since issuance of our report to the Committee, we have had further j

l discussion with-the appidcant on the foregoing r.atters.

In addition,_

i further inferr.atien on some of these matters was provided in Amendments

es. 23 and 24.

The additional information has been reviewed and our evaluatien is presented in this report.

Our review of Technical Specifications is continuing and a revised draft proposed by the applicent (/cendment 23) has been provided to the-Corr.ittee.

Eubject to resolution of items 1, 4 and 6a above, development

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of final Technicc3 Specifications, and satisf actory completion-of con-struction and preoperational tests, we conclude that the Monticello plant may be licensed for operation.

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' 2.0 DISCUSSION 2.1 Non-Isolable Break in the HPCI Steam Supply Line The applicant has not formally submitted am additional information-t in regard to this matter.

However, he has informed us orally that a mdification vill be made to reduce the spectrum of break aires ove.r which a break in the. F CM steam supply line, occurring in the main steamline vill not initiate automatic isolation of the broken line. We understand that this will be accomplished L replacing-the pressure taps in the HPCIS steamline with a more sensitive orifice device to measure an increase in steam flow above rated, which would be indicative 'of a break in the itPCI steam 31ne. Also, the applicant.has stated analyses indicate that, in the event of a break which cannot be detected immediately, sufficient time is availabic to assess an abnormal condition and then manually actuate the auto-relief system without resulting in any significant offsite doses.

As stated in our previous-repert-to the Committee,'our position is.

that the applicant will need to (a) demonstrate that the HPCIS and RCICS steam supply lines have adequate l isolation capability with the present design, or (b) present information on changes that could be made'so that.

failures in these steamlines could be detected, and-the lines isolated-as required.

The applicant plans-to discuss.this matter =vith us prior to the ACRS meeting.

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4. i cestings were made. Because of these deficiencies and the record of failure ins-1vio these components, we still believe that additional inservice inspection of thue castings should be required <

k'e intend to include in the Technical Specifications an inspection program for the remaining carburized castings which will include inspection of all castings followinp, the hot functional test and after the first refueling shutdown, and a lower frequency of inspection thereafter.

k'e have orally informed tha applicant of our intentions. At this time the applicant has not made a commitment to accept this prooosed inservice inspection program.

2.4 Vibration Monitorine of Reactor Internals The applicant has not changed his position from that previously.

reported to the Committee, i.e.,

for the Monticello plant, Northern States Power will rely upon the results of the Dresden 2. vibration tests.

The applicant stated that the results of the Dresden 2 vibration tests will adequately represent any vibration _that may occur in the Monticello reactor, since the fluid velocities within the Dresden 2 vessel are expected to be higher than in Monticello.

As stated in our previous report, we plan to require that the vibration. levels of the critical internal'and recirculation system components be monitored during plant startup and the initial operation period.

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The applicant has stated that he is prepared to discuss this matter in detail with the Committet t the January meeting.

2.5 Seismic Design of Structures Since the submittal of our report to the Committee, additional information on stress levcis in critical piping systems and discussions on analytical methods employed for dynamic analysis have been provided by the applicant in Amendment 24, dated December 19, 1969. This informa-tien is being evaluated by cur seismic design consultant, and his report.

will be submitted to the ACES before the full Committee meeting. We 9

anticipate that the seismic design methods employed will be accepteble.

2.6 Instrumentatien ar.d Control Items 2.6.1 ECCS Low-pressure Auto-relief Interlock The interlock function is provided by an arrangement involving six e

pressure switches (ene switch for each ECCS pump) monitoring pump discharge pressure.

The interlock is a permissive-type which allows initiation of auto-relief chen one of the six switches responds-to either the pressure characteristic or to a circuit failure which produces the same signal to the auto-relief system as the pressure characteristic. Therefore, a single failure which produces a permissive condition would defeat the purpose of the interlock in cases when non permissive circuit orientation is required. We have concluded that this' is not acceptable and that single failure immunity is required.

The applicant does not agree to this requirement.

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j 2.6.2 Monitoring Capability of the Integrity of Bellows on the Target Rock:

Relief /Saft + Valves Because the integrity of.the bellows is essential to the self-actuated operating mode of the Target-Rock relief / safety valves, we

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conclude that the testing circuits associated with bellows integrity monitoring should be improved to permit unambiguous testing during operation.

The applicant statad that such capability would be provided in the.

Monticello design and we conclude this is acenptable. We shall review-the specific design details prior to issuance of an operating license to ensure that the< - requirements are satisfied.

2.6.3 Acceptab'11ty ef Standby Gas Treatment System (SGTS) Initiating Circuitry Functional.1y, the SGTS is designed such that one (filter) train is considered to be the " preferred" system and is actuated initially.

If' this "1.ref erred" train f ails, - the redundant filter train is actuated after a time delay of 50 seconds.

Since submittal of.-our report to the Committee, we have received and revieced the elementary diagrams for the SGTS. The applicant has stated l

tha; the system is designed to those portions of IEEE.279 which relate -

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to testability and the single failure criterion. We conclude that-the system is comprised of two electrically independent and redundant initia-

• ing systems and have not uncovered any deficiencies.

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In response to our questioning concerning the ef fect on of fsit '

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this initiation circuitry will be changed. Upon receipt of the initiation signal, both filter trains will be started simultaneously; when opera-bility is assured one train will be manually shutdown.-

We consider this acceptabic.

2.6.4 Single Failure Criterion At the time we wrote our report to the Committee, some ambiguity existed as to whether the reactor protection and containment isolation I

sys tems were designed to meet the single failure criterion as defined in IEEE 279.

In Amendment 23 the applicant confirmed that the systems are so designed. This matter is now resolved.

2.6.5 Seismic Testine of Class I Instrumentation In our earlier report to the Committee, we stated that the-applicant had not subaitted a completion schedule for the seismic test program related to the balance-of-plant systems ; i.e., other than General Electric-supplied systems.

In Amendment 17 the applicant stated that the test program for the General Electric-supplied instrumentation similar to that to be installed at Monticello will be completed by December 31, 1969.

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• 9 of function.. A program is currently underway-to ascertain how the equipment inanuf acturers are able to assure-compliance with that portion of the specifications. The applicant stated that it.is currently expected I

that this effort will be completed by March 15, 1970.

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