NRC-90-0060, Responds to 900104 Request for Supplemental Station Blackout Submittal

Responds to 900104 Request for Supplemental Station Blackout Submittal

ML20042D747
Person / Time
Site:	Fermi
Issue date:	03/29/1990
From:	Sylvia B DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-90-0060, CON-NRC-90-60 NUDOCS 9004050299
Download: ML20042D747 (3)

Text

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I March 29, 1990 1

NRC-90-0060 j

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I i U. S. Nuclear Regulatory Commission  ;

l Attn Document Control Desk i Washington, D. C. 20555 l

Fermi 2 i

References:

1) i j NRC Docket No. 50-341 NRC License No. NPF-43

2) Detroit Edison Station Blackout Submittal to NRC '

NRC-89-006), dated April 17, 1989 NUMARC Letter " Station Blackout (590) Implementation

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3)  !

Request for Supplemental $80 Fubmittal to NRC" dated 1 January 4, 1990 l

Subject:

Detroit Edison Response to Request  ;

for Supplemental SB0 Submittal to NRC  ;

Detroit Edison has performed a review of its station blackout submittal J (Reference 2) as requested by WUNARC (Reference 3). The results of our t review are addressed in this letter. The review was performed to address NRC concerns with proper documentation and consistent implementation of l NUMARC 87-00 guidance. We also reviewed our previous station blackout l submittal in light of the clarification provided by WUNARC in their January l 1 4, 1990 letter (Reference 3). Where appropriate, we have addressed these  ;

i clarification items in this submittal, j i

f Our review affirms that the previous Detroit Edison station blackout  ;

i submittal (Reference 2) was based on use of NUMARC 87-00 guidance. The applicable NUMARC 87-00 assumptions are documented and are on file. Where there have been departures from NUMARC 87-00 methodology, they have been l documented and are discussed in this submittal, j The review of the station blackout submittal affirms that no hardware ,

changes to the plant will be required to meet the station blackout rule. j l

In its station blackout submittal Detroit Edison committed to perform a .

battery capacity test of the Non-Class IE BOP-2PC 260/130 volt battery. l The battery capacity test was completed during the first refueling outage i and the BOP battery performed satisfactorily. The required battery .

l capacity to support station blackout operation is 100% of nominal.

Therefore, battery replacement is not necessary.

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l f USNRC March 29, 1990  !

l NRC-90-0060 [

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l Procedure revisions needed to meet NUMARC 87-00 will be completed by April l j

- 14, 1990, as per the original submittal. l l An EDO Reliability Program similar to that required by NUMARC 87-00 has been in place for some time at Fermi 2. EDG st, arts and failures are  !

l tracked by procedure, and the target 0.95 availability has consistently

, been exceeded. Note that the Fermi 2 Technical Specifications require a [

similar tracking program.

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0 j NUMARC 87-00 assumes that the control room does not exceed 120 F during a  !

station blackout and that the primary containment temperature during the  !

j event is enveloped by LOCA profiles. Detroit Edison has verified NUMARC's l

, assumption that the Fermi 2 control room would not exceed 1200 F during a  :

J postulated blackout event. Because Ferai 2's control room has a variety of i

construction materials that must be considered, a unique compart.nent heat ,

up analysis was performed that. takes credit for other types of wall  ;

I materials (plaster, block walls, etc.) and takes credit for the volume  !

! above t.he partial suspended ceiling within the control room. This analysis  !

is documented in a design calculation on file at Fermi.

Detroit Edison also performed a plant specific heat up analysis of the  !

primary containment. The analysis does not assume reactor depressurization l and uses the drywell atmosphere, cont.ainment steel liner and sacrificial j shield steel liner as heat sinks. The analysis was redone to account for  ;

an assumed 25 spa seal leakage from the reactor recirculation pumps. This i is a conservative assumption because 25 spa is the total react.or coolant j system leakage when averaged over any 24-hour period that is allowed by the  ;

plant Technical Specifications. The analysis is documented in a design '

calculation and concludes that the cont.ainment design temperature of  :

3400 F is not exceeded within the first hour of station blackout,. The i drywell temperatures given in Det.roit Edison's initial st.ation blackout j submittal have been revised to include the assumption of reactor i I

recirculation pump seal leakage.

Detroit Edison's original response regarding compressed air usage remains ,

unchanged. However, the statement t. hat no air-operated valves are relied '

upon to cope with a station blackout for one hour needs to be clarified. i As noted in the response, the two low-low set relief valves and five ADS i t safety relief valves rely on nitrogen to operate and each valve has an  ;

accumulator sized to provide five actuations each. These valves will be  !

available and are relied upon to cope with a station blackout.  :

A final clarification regarding cont.ainment isolation from the original  !

station blackout submittal is appropriate. Valve position verification  ;

I will be determined by visual observation of valve stem position at the i valve location or by control panel indicating lights as available. ,

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USNRC Harch 29, 190 NRC-90-0050 Page 3 If there are any quest,1ons regarding this submittal, please contact Mr.

Lewis Bregni at (313) 586-4072.

Sincerely, h'fctt .. A' cc: A. B. Davis R. W. DeFayette W. G. Rogers J. F. Stang n

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