ML20245J250

Provides Clarification Re Proposed Amend to Tech Specs Section 3.15, Auxiliary Electric Power Sys. Areas That Warrant Clarification Can Be Grouped Into Impact or Risk Assessment of Diesel Generator & Use of LOCA Events

ML20245J250
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	08/04/1989
From:	Trzyna G COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8908170597
Download: ML20245J250 (2)

Text

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m Commonwealth Edison

. 72 West Adams Street, Chicago, liknois

' * , A~ddress Reply to: FoMic~e Efo'T/B7-Chicago, lihnois 60690 - 0767 August 4, 1989 Director of Nuclear Reactor Regulation US Nuclear Regulatory Commission Mail Station P1-137 Hashington, DC 20555

Subject:

Zion Nuclear Power Station, Units 1 and 2 Proposed Amendment to Technical Specifications Section 3.15, " Auxiliary Electric Power System" License Nos. DPR-39 and DPR-48 MRC Docket Nos. 50-295 and 50-304

Reference:

July 6, 1989 letter from GE Trzyna to US NRC/NRR Gentlemen:

Our submittal dated July 6, 1989 provided data to support the proposed Technical Specification amendment which would allow a 45 day outage of the "0" Diesel Generator (DG) during the forthcoming Unit 1 Refueling outage. Subsequent to your review of our submittal, several telephone conference calls between CECO /NRR personnel indicated the need for clarification to some items of our submittal. The purpose of this letter is to provide the necessary clarification.

The areas that warrant further clarification can be grouped into two broad categories:

1. Clarify the impact or risk assessment of the "0" Diesel Generator on Unit 1, ie, the shutdown unit; and

2. Clarify the use of only LOCA events in your PRA assessment and the selection of hot leg versus cold leg break scenarios.

Recardina Cateaory 1: (Note: The dates provided are meant to be illustrative and may change, but the sequence is not affected.)

1. Unit I will be taken off-line on September 8, 1989.

2. "0" Diesel Generator will be taken out of service on September 10, 1989 following a successful test of 1A and 1B DG's.

3. The entire core will be off-loaded from September 15, through October 25, 1989.

4. The 45 day out-of-service clock for "0" DG ends on October 25, 1989.

5. The Residual Heat Removal (RHR) pumps are not powered from "0" DG.

6. No mid loop operations are planned for this outage while fuel is in the vessel.

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.Dir. Of Nuclear Reactor Reg. - I2 - August 4, 1989-

c.I. 7. PRA analysis indicates that the "0" DG does.not at all. impact the likelihood of core melt.

The conclusion drawn from an evaluation of these variables indicates

'that there-are two 5-day periods where "0" DG will be out of service and.where' fuel will.be present in the vessel. This is.within the current frequency of; the allowable DG outage time of 7. days, even while operating, provided no other electrical failures are evidenced. The risk assessment is further minimized when consideration is'given to the fact that there will not.be any fuel in the vessel for 30 to 35 days of the OG outage. Our proposed Technical Specification does provide-guidance to address ECCS pump failures or another DG failure on either the operating or the shutdown unit.

Recardino Cateaorv 2:

Question No. 1:

Why did the submittal; consider only LOCA' events and not consider

. transient events given the relative frequency of the two types of initiating events in the ZPSS?

Answer: It was understood that the primary thrust of the NRC's interest was the LOCA originated events. However, even though the transient initiated events have a higher initiating event frequency  ;

than LOCA events, core damage for transient initiated events requires the failure of the entire, three train auxiliary feedwater system in addition to the failure of appropriate ECCS trains. This tends to lower the probability of such events significantly. If, for exaltple, one does lose aux. feedwater, the core can still be cooled by feed and bleed using the high head ECCS (1/4 pumps) and the pressurizer PORV's. The fact that transient events are not the dominant core melt sequences for the ZPSS tends to demonstrate this conclusion.

Question No. 2:

The MAAP analysis employe the hot leg break as the basis for the l study enclosed. What difference would result from the use of the cold leg as a break location?

I Answer: Experience with the MAAP code indicates that no significant difference would result in terms of the long term response (hours) of the plant or possible recovery actions. More importantly, the role of the MAAP analyses in this submittal is one of confirming the 1/4 high head ECCS pump success criteria used in the ZPSS for the small '

break. Use of the cold leg versus hot leg break in the analyses '

would not, in any event, affect the calculated probabilities in this submittal since questions of timing and long range recovery are not carried into these probabilistic evaluations.

Please direct any questions that you may have to this office.

Ver truly yours,

'#M ,

g4 G.E. Trzyna Nuclear Licensing Administrator  ;

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/sc1:0242T:1-2 j cc: Chandu Patel-NRR I