ML20212E650

Forwards Request for Addl Info Re Confirmatory Issue 49, Tracking Remaining Backfit 4 Items Concerning Steam Generator Control & Protection,Per 850509 Meeting & Recent Correspondence.Prompt Response Requested

ML20212E650
Person / Time
Site:	Beaver Valley
Issue date:	02/19/1987
From:	Tam P Office of Nuclear Reactor Regulation
To:	Carey J DUQUESNE LIGHT CO.
Shared Package
ML20212E636	List: ML20212E632 ML20212E650
References
TAC-62936, NUDOCS 8703040402
Download: ML20212E650 (7)

Text

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February 19, 1987

' Docket No. 50-412 Mr. J. J. Carey, Senior Vice. President Duquesne Light Company Nuclear Group Post 0ffice Box 4 Shippingport, PA 15077

Subject:

. Beaver Valley Unit 2 - Transmittal of Staff Position On :

Steam Generator High Level Trip During increased Feedwater Flow Event (TAC 62936)

Confirmatory Issue 49 was opened to track the actions that remained from Back-fit Issue 4, which concerned steam generator level control and protection. As a result of the appeal meeting of May 9,1985, and recent correspondence be-tween Duquesne Light Company and the NRC, we believe additional effort on your part is needed.

The enclosed evaluation documents the background, current status and the infonnation needed to resolve the issue. We request your prompt response to this matter.

This information request affects fewer than 10 respondents; therefore, OMB clearance is not required ttnder PL 96-511.

Sincerely, Peter S. Tam, Project Manager PWR Project Directorate #2 Division of PWR Licensing-A

Enclosure:

AS stated cc w/ enclosure:

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I l Mr.' J. ;J. Carey Duquesne Light Company-Beaver Valley ? Power Station CC*

Gerald Charnoff, Esq.

Mr. R. E. Martin, Manager l Jay E. Silberg, Esq.

Regulatory Affairs 3

Shaw, Pittman, Potts A Trowbridge Ducuesne Light Company l2300 N Street, N.W.

Peaver Valley Two Pro.iect Washinoton, DC 20037 P. O. Box 328 Shippingport, Pennsylvania 15077 Mr. C. W. Ewina, Quality Assurance Zori Ferkin Manager-Assistant Counsel Ovality Assurance Department Governor Energy Council Duquesne Light Company 1625 N. Front Street

-P. O. Box 186 Harrisburg, PA 15105 Shippingport, Pennsylvania 15077 John D. Burrows, P.E.

Director, Pennsylvania Emergency Director of Utilities Management Agency State of Ohio Room B-151 Public Utilities Commission Transportation 8 Safety Building _

180. East Broad Street l

Harrisburg, Pennsylvania 17170 Columbus, Ohio A3266-0573 Mr. T. J. Lex Bureau of Radiation Protection j-Westinghouse Electric Corporation PA Department of Environmental Power Systems

. Resources

- P. O. Box 355 ATTN:

R. Janati Pittsburgh, Pennsylvania 15?30 P.O. Box 2063 Harrisburg, Pennsylvania 17170

-Mr.'P. RaySircar Stone & Webster Engineering Corporation BVPS-2 Records Management Supervisor P. O. Box 2325' Duquesne Light Company Boston, Massachusetts 02107 Post Office Box 4 Shippinoport, Pennsylvania 15077 Mr. J. Beall l

-U. S. NRC John A. Lee, Esq.

j' P. O. 181 Duouesne Light Company Shippingport, Pennsylvania 15077 10xford Centre 301 Grant Street Mr. Thomas E. Murley, Regional Admin.

Pittsburgh, Pennsylvania 15279 i

U. S. NRC, Region I 4

631 Park Avenue King of Prussia, Pennsylvania 15229 i

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ENCLOSURE

.NRC POSITION ON REAVER VALLEY 2 STEAM GENERATOR HIGP LEVEL TRIP.

FOR INCREASED FEEDWATER FLOW ACCIDENT 1.

BACKGR0 LWD This issue criainated at least two years ago when the Nuclear Reculatory Connission (NRC) staff enforced a control and protection system recuirement as it applied to the steam generator water level channel instrumentation ir

Peaver Valley Power Station Unit 2 (BVPS-2). The NRC staff had found that this instrumentation system was part of the Peactor Protection System (RPSI and yet was not in compliance with IEEE Standard 279-1971, as required under r

10 CFR 50.55a(h). Duquesne Light Company (DLC) took the position that the steam generator high level trip was not part of the RPS and therefore did not need to comply with IEEE Standard 271CT971.

DLC appealed the NRC position to the Division of Licensing. On May 9, 1985, E

an Appeal Meeting was held (see ref. 31 Soon afterwards Hugh L. Thompson, who presided at the Appeal Meeting, wrote a letter to DLC (see ref. 51 stating, "Before this issue can be resolved, an additional review of the 4

applicability of 10 CFR 50.55a(h) to this issue must be completed," and also stating, "I need to have your Final Safety Analysis Report (FSAR) revised to eliminate the inconsistency in representing the steam generator high level trip system as part of the Emergency Safety Features Actuation System."

II. NRC POSITION i

- The NRC staff's position depends on whether or not the trip signal is part of the Emergency Safety Features Actuation System (ESFAS). The Appeal Meeting seems to have ruled that the steam generator high level trip is not part of ESFAS (the record is not perfectly clear; see page 32 of the transcript,ref. 3). It is therefore necessary to restate that staff's position with regard to this issue.

In order to assure that the DLC accident analysis is valid for the increased 3

feedwater flow event due to feedwater malfunction, the following criteria must be met:

1.

A condition II event, by definition, cannot be allowed to graduate into a condition III or condition IV event, 2.

Analysis must be provided until the time when the reactor has stabilized (analysis shall not be terminated while the steam generator is being filled),

-and a

3.

The event must be mitigated with safety-related systems or operator action if sufficient time is available (since the high level trip is not part of ESFAS, this trip cannot be relied upon to perform any ESFAS function).

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.;. i III. DLC PROPOSAL Currently, the Beaver Valley Unit 2 Final Safety Analysis Report takes credit for the high steam generator level as a trip function assumed operable in the accident analyses. This trip is referenced in 15.1.F. as the mitigating signal for the increased feedwater flow event due to feedwater malfunction.

It is listed in Table 15.0-4:

" Trip Points and Time Delays to Trip Assumed in Acci-dent Analysis." It is also assumed in the analysis using LOFTRAN referenced-

.by Table 15.0-3 and Figures 15.1-1 throuch 15.1-2a.

DLC's 'first proposal (see ref. 61 to eliminate the FSAR inconsistencies was simply to transfer the trip signal from.the "ESFAS" column of Table 15.0-6 to the "0THER" column. Since the inconsistencies are found throughout the FSAR, this change to a single table in the accident analysis did not resolve the problem.

A second DLC proposal was submitted (see ref. 7) based on three main points.

Probabilistic risk assessment (PRA) demonstrated the unlikelihood of the event under concern. Operator corrective action was proposed to be taken in less time than the staff-specified guidance of 10 minutes for overfill events at low power.

It was also suggested that the staff's position was a backfit.

DLC's PRA analysis evaluated the probebility of failure of the steam genentor

-high level trip during an increased feedwater flow event. However, PRAs alone do not constitute a justifiable basis for NRC licensing actions related to ESFAS and Chapter 15 accident analyses, the review of which is primarily deter-ministic and defined in the Standard Review Plan.

The backfit rules do not apply (see ref. 3, p.8 and ref. 41 since the i

requirement is not new and has been applied since issuance of 10 CFR 50.55a(h).

The Steam Generator Hi-Hi Level Trip in Chapter 15 of the BVPS-2 FSAR has been interpreted by NRC to be part of a safety-related system. If DLC differs with this position it must substantiate and make the appropriate changes to the FSAR.

DLC proposes that credit for operator action be given for the increased feedwater flow event.

In reference 2, the Westinghouse-supplied analysis shows that the time available before the filling of the steam generator is greater than 10 minutes at full power, both for beginning of core power and end of core life.

Recently though, the staff has received a submittal from DLC (ref. 9) reanalyzing the increased feedwater flow accident. This latest evaluation indicates that the staff-specified guidance of 10 minutes for operator corrective action is not met at full power. Overfill of the steam generator, previously estimated at 11.7 minutes, is now approximated at 6.5 minutes. However, even if the operator had 10 minutes to act, the mo're limiting case would be at low power.

Therefore, analysis would need to be tirovided to show that operator action was creditable in the low power case as w411.

Also listed in reference 2 is an alarm sequence during a feedwater control malfunction. The table consists of a number of alanns and annunciators relied upon to alert the operator to the increased feedflow condition. However, those alarms listed are actuated by the steam generator level transmitter

3-which should be assumed to be unavailable since the equipment is not safety-related. Alams must be furnished to ensure that the operator will be warned of steam generator overfill.

Also, in references 2.and 9, DLC concludes that, "The results for the first scenario show that feedwater addition to affected steam generator will be terminated on coincidence of P-4 and Low RCS T"Y6) prior to steam generator i

3 overfill." Since DLC has considered (see ref

. that low T coincident withP-4isnotneededforsafety,anynewanalysisprovided6V9y 0LC must not utilize this coincident signal to teminate a Chapter 15 event (see Section II above).

It has been stated in references 3 (pp.10 and 23) and 8 that the high steam generator level function has been typically included as a non-safety-related system for over 25 other Westinghouse-NSSS plants found acceptable by NRC.

This statement is misleading. Those plants reviewed and approved by the staff differ from BVPS-2 in that:

they had additional safety-related or non-safety-related channels, they used a completely different approach, such as crediting the safety injection system for mitigation of the event, they considered the steam generator high level trip to be safety-related and part of the ESFAS (Beaver Valley, Unit I fits in this category),or they were reviewed before the IEEE requirements were in effect.

And, finally, DLC sent the staff a third proposal (see ref. 8) which changes the FSAR wording of Chapter 15. Most of the FSAR pages that reference the use of the steam generator high level trip to teminate the event have been changed into words that allow use of the trip without relying on its operability. However, they do not cite the teminating event for the accident and do not provide the appropriate revisions to the LOFTRAN analysis, which now assumes that the steam generator high level trip mitigates the event.

DLC states that the trip is a " convenient ending point for the analyses" (ref. 3, pp. 17 and 18, and refs. 6 and 8). This is not appropriate; a j

legitimate stopping place must be defined, or analysis be provided to show that the three criteria in Section II of this report are met.

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IV. RESOLUTION Resolution of this issue may be attained in one of three ways.

I i

First, DLC may request an exemption from 10 CFR 50.55a(h) which requires j

compliance with IEEE Standard 279-1971,~ paragraph 4.7.3.

This would l

reestablish the trip signal as safety-related thereby eliminating the need for FSAR changes, and only lift the requirements of IEEE Standard 279-1971 for the trip signal derived from the increased feedwater flow event.

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mentation and controls for each steam generator to bring the design into con-formance with IEEE Standard 279-1971:

(1) add a fourth safety-grade level

. channel and convert'the high level trip logic from 2 out of 3 to 2 out of 4;

-(2) eliminate the present safety-grade level signal input to the level control system and add a fourth (non-safety gradei level channel for level control,

. while retaining the present ? out of 3 high level trip; or (3) add a median selector device (non-safety grade) which would choose the middle si

.the.three existing level channels (through safety-grade isolators) gnal from for input to the level' control, while retaining the present 2 out of 3 high level trip.

And third. DLC may reanalyze the accident analysis of the increased feedwater flow event without credit.to.the steam generator.high level trip for terminating the event, in compliance with reference 2 which cites the decision of the Appeal Meeting.

In this ' case, the applicant uill verify that the trip signal is not part of the reactor protection system because it is not safety-related. Therefore IEEE Standard 279-1971 would not apply. This would be accomplished by the reanalysis specified in Section II of this position.

If the applicant wishes to resolve this issue by eliminating FSAR inconsisten-

.cies in the accident analysis (Chapter 15) but wishes to use a different 1

approach, such as crediting operator action, then the applicant may submit such a proposal for review provided that this proposal meets the criteria specified in Section II of this report.

V.

REFERENCES 1.

Beaver Valley Power Station, Unit 2, Final Safety Analysis Report, Chapter 15.

2.

Letter from E. J. Woolever, DLC, to G. W. Knighton, NRC, dated June 8, 1984.

3.

U.S. NRC Proceedings: " Reaver Valley - Unit 2 Packfit Appeal Meeting on Steam Generator Water Level Channel Instrumentation Issue," Thursday, May 9, 1985.

4 Memorandum, R. Bernero (NRC) to H. L. Thompson (NRC) dated June 15, 1985 (this internal memorandum has been placed in the public document room).

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5.

Letter from Hugh L. Thompson, NRC, to John J. Carey, Duquesne Light Company, dated November ??, 1985.

6.

Letter from John J. Carey, DLC, to Hugh L. Thompson NRC, dated December 20, 1985.

7.

Letter from John J. Carey, DLC, to Narold R. Denton, NRC, dated March 10, 1986.

8.

Letter from John J. Carey, DLC, to Harold R. Denton, NRC, dated August 19, 1986.

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9.

. Letter from John J. Carey, DLC, to Harold R. Denton, NRC, dated January 15, 1987.

10 Beaver Valley Power Station, finit 2. Safety Evaluation Report, NUREG-1057, Section 7.3.3.7, October 1985.

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