ML19341B294

Responds to NRC Safety Evaluation Requiring Installation of Addl 100% Capacity Auxiliary Feedwater Sys.Proposes Risk Reduction Comparison to Provide Evaluation of Acceptable Alternatives

ML19341B294
Person / Time
Site:	Davis Besse
Issue date:	01/23/1981
From:	Crouse R TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
677, TAC-43516, NUDOCS 8101300605
Download: ML19341B294 (3)

Text

Docket No. 50-346

" UU License No. NPF-3 Serial No. f,77 EDISON January 23, 1981 P c>.wascan u e %ert

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mau w Director of Nuclear Reactor Regulation Attention:

Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Licensing United States Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Reid:

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on July 6, 1979, the NRC issued a letter lifting the Confirmatory Order dated May 16, 1979, allowing the Davis-Besse Nuclear Power Station Unit 1 (DB-1) to return to power following the Three Mile Island Unit 2 accident. The safety evaluation attached to that letter indicated that the NRC would at some future time require the installation of an additional 100 per cent capacity auxiliary feedwater pump at Davis-Besse Unit 1.

Since July of 1979, Toledo Edison has been evaluating some options to meet our understanding of the NRC requirements of such a system modification. A review of past actions is appropriate to fully understand Toledo Edison's current activity.

Davis-Besse Unit I was put into operation with a fully safety grade auxiliary feedwater (AFW) system. The Final Safety Analysis Report (FSAR) reflected this design. At that time, the system was unique among its Babcock & Wilcox (B&W) nuclear steam supply system (NSSS) counterparts due to the full extension of safety grade criteria and requirements to include not only the mechanical systems but the instrumentation and control systems as well. Davis-Besse Unit I has installed in its original design an AFW initiation system (the Steam and Feedwater Rupture Control System) and a steam generator level control system, both completely independent of the B&W supplied integrated control system (ICS).

The basic criteria of these systems was to isolate the steam generators, pro-vide auxiliary feedwater and control level within 40 seconds of initiation. The 3-1 Technical Specifications Section 3/4-3.2 reflects these requirements. When put in operation the DB-1 AFW system complied in all respects to the NRC safety grade requirements.

In the NRC's Safety Evaluation Report, an additional requirement that one train of AFW would have to be independent from AC power requirements was imposed.

This later requirement was identified as a license condition for modification that has been met by system modifications during the recent refueling outage.

These requirements were met using two safety grade steam driven pumps.

Following the accident at Three Mile Island on March 28, 1979, many questions were directed toward AW systems at B&W NSSS units. NRC concerns about DB-1 fool 3

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Docket No. 50-346 License No. NPF-3 Serial No.

677 January 23, 1981 AFW system seemed to concentrate on:

1.

Overall system reliability.

2.

Potential loss of AFW pump motive power.

i. Steam generator cooling requirements to reduce reactor coolant system pressure after small break loss of coolant accidents and after complete loss of feedwater transients.

In responses to the staff, each of these items were discussed. Toledo Edison filed submittals on Item 1 above on May 23 and July 3,1979, as well as Topical Report BAW 1584, " Auxiliary Feedwater System Reliability Analyses -

A Generic Report for Plants With Babcock & Wilcox Reactors." Additionally, as reported since these submittals, Toledo Edison completed its license con-dition, in response to Item 2 above, by diversifying power supplies to the auxiliary feedwater motor-operated valves. This now insures a train operable without alternating electrical current available.

On the potential loss of motive power, the staf f utilized a non mechanistic, complete and instantaneous loss of steam pressure in both steam generators (SG) as its basis of concern. 'ihin ignored completely the safety grade systems that provide SG isolation and AFW initiation. Toledo Edison's submittal on June 23, 1979, illustrated by calculation and experience the ability of the design to respond appropriately to reduced pressure auxiliary feedwater pump turbine starts. To Mrther compound these assumed multiple safety system failures, the NRC staff would not consider any capability to re-estabilsh steam pressure as motive power to the AFW pumps. On May 22, 1979, Toledo Edison's submittal provided B&W's " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant - Volume III - Raised Loop Plant."

This illustrated the additional time delay available for such action (greater then 30 minutes) that the DB-1 raised-loop reactor coolant system design would be able to sustain during a complete loss of all feedwater, regardless of the source.

By ignoring the unit's design capabilities, the assumptions imposed by the NRC staff predetermined their conclusion that an additional 100 per cent diverse power capability need be provided.

The third NRC concern related to the reliance on the safety grade AFW system to depressurize the reactor coolant system to e:ithin the effective pressure range cf the DB-1 high pressure injection pumps. As indicated above, the analysis identified that greater than 30 minutes was required without feedwater from any source to provide a potential problem. This means, given an event the unit can recover without any core damage as long a, 550 gpm of any source of feed-water is provided to a steam generator within at least 30 minutes of initial total feedwater isolation. Additionally, the same submittal provided descrip-tions to depressurize the reactor coolant system even with such a delay.

These scenarios have been factored into our operating procedures.

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Docket No. 50-346 License No. !!PF-3 r

serisi No. 677 4

i January 23, 1981 1

Your July 6, 1979, letter indicated regardless of information submitted, an intended requirement was to add an additional 100 per cent capacity AFW pump.

l Starting from this NRC evaluation, Toledo Edison has proceeded on a feasi-bility study evaluating design options of such a backfit to DB-1.

The study was undertaken with criteria preceived by us as adequately addressing the NRC staff's basic concerns. The results are in each case extremely i

costly and require long lead times, reflecting major plant additions and/or i

modifications as well as site alterations.

In addition, there are AFW system operational philo.tophy differences compared to the present system.

In re-

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viewing the NRC intended purpose for such a modification, and relating it to the magnitude of the physical change required, Toledo Edison has decided to undertake an attempt to quantify the relative risk reduction actually pro-vided by such a modification.

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To bring this issue to final resolution, it is proposed that, prior to pro-ceeding on any major plant modification, a risk reduction comparison be completed to provide an evaluation of the acceptable alternatives. This 4

would allow us to optimize the plant response results, minimize the plant i

perturbations and still verify that the design provides an appropriate level i

of protection to the public health and safety now and af ter any such modifi-i cation is complete. A meeting is proposed to identify the pe-formance criteria i

and risk reduction results behind the proposed modification on-the July 6, 1979, j

NRC letter. We expect this information to be available for discussion in February, 1981, i

j Very truly yours,

/CL _

RPC:TJ!!:aa cc: DB-l NRC Resident Inspector

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