ML13331B423

Provides Response to IE Bulletin 80-04,Item 2,per 800716 Commitment.Westinghouse Reviewed Previous Analysis of Core Response Following Main Steam Line Break.Results Indicated No Main or Auxiliary Feedwater Assumed in Previous Analysis

ML13331B423
Person / Time
Site:	San Onofre
Issue date:	08/04/1980
From:	Ottoson H SOUTHERN CALIFORNIA EDISON CO.
To:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
IEB-80-04, IEB-80-4, TAC-44652, NUDOCS 8008120377
Download: ML13331B423 (3)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8008120377 DOC.DATE: 80/08/04 NOTARIZED: NO DOCKET #

FACIL:50-206 San Onofre Nuclear Station, Unit 1, Southern Californ 05000206 AUTH.NAME AUTHOR AFFILIATION OTTOSONH.L.

Southern California Edison Co.

RECIP.NAME RECIPIENT AFFILIATION ENGELKEN,R.H.

Region 5, San Francisco, Office of the Director

SUBJECT:

Provides resoonse to IE Bulletin 80-04,Item 2,oer 8,00716 commitment.Westjnahouse reviewed previo.us analysis of core resconse following main steam line break.Results indicated no main or auxiliary feedwater assumed in previous analysis.

DISTRIBUTION CODE: 4001S COPIES RECEIVED:LTR J ENCL 2

SIZE:

TITLE: General Distribution for after Issuance of Operating Lic NOTES:1 coov:SEP Sect. Ldr.

05000206 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL ACTION:

CRUTCHFIELD 13 1

INTERNAL: D/DIR,HUM FAC S 1

DIRHUM FAC SFY 1

I&E 12 2

NRC POR 02 1

OE D 114 1

)

OR ASSESS BR 19 1

E G FI 01 1,

EXTERNAL: ACRS 20 16 1

LPDR 03 1

NSIC 04 1

AUGTN O E U LE TOTAL NUMBER OF COPIES REQUIRED: LTTR ENCL

Southern California Edison Company 0

BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD CALIFORNIA 91770 August 4, 1980 U. S. Nuclear Regulatory Commission Attention:

R. H. Engelken, Director Office of Inspection and Enforcement Region V Suite 202, Walnut Creek Plaza 1990 North California Boulevard Walnut Creek, California 94596 Gentlemen:

Subject:

Docket No. 50-206 IE Bulletin 80-04, Analysis of a PWR Main Steam Line Break With Continued Feedwater Addition San Onofre Nuclear Generating Station Unit 1 By letter dated July 16, 1980 we advised you that we would complete our.review and provide our response to Item 2 of IE Bulletin 80-04 by August 1, 1980. The purpose of this letter is to provide that response.

Item 2 of IE Bulletin 80-04 requested licensees to review analysis of the reactivity increase which results from a main steam line break inside or outside containment to determine if previous analysis considered all potential water sources and if the reactivity increase is greater than previous analysis indicated.

In response to our request, Westinghouse reviewed the previous analysis of core response following a main steam line break for San Onofre Unit 1. The results of the review showed that no main or auxiliary feedwater had been assumed in the previous analysis.

Subsequently, Westinghouse performed a reanalysis of this event. The cases reanalyzed were a main steam line break (complete severance of a pipe) outside contain ment at no load conditions with offsite power available, and an 80o0 800812o 3"

Mr. R. H. Engelken August 4, 1980 accidental depressurization of the main steam system associated with the inadvertent opening of a single steam dump, relief, or safety valve with offsite power available. These cases conserva tively assumed main feedwater flow addition until main feedwater isolation on the safety injection signal and auxiliary feedwater runout flow initiated coincident with the event. The results of the reanalysis confirmed that the main steam line break transient results for these cases are very insensitive to continued feed water addition for San Onofre Unit 1. It is expected that the results for other no load cases previously analyzed and full load cases (previously shown to be less limiting) would also be insensitive to continued feedwater addition based on Westinghouse generic studies.

The first minute of the transient is dominated entirely by the steam flow contribution to.primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core. The effect of auxiliary feedwater is minimal. The primary side pressure, on which the low pressurizer pressure safety injection signal is based, decays at a slightly faster rate with the addition of auxiliary feedwater. This accelerates the safety injection signal actuation

(<.5 second sooner) as well as allowing a slightly greater safety injection flowrate with the faster pressure decay. These two effects compensate for the increased cooldown rate. The overall results are, therefore, negligibly impacted with the.addition of auxiliary feedwater flow.

The auxiliary feedwater flow becomes a dominant factor in determining the duration and magnitude of the steam flow transient during later stages in the transient. However, the limiting portion of the transient occurs during the first minute, both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system.

Hence, the conclusions documented in the previously submitted main steam line break core response analysis for San Onofre Unit 1 remain valid and applicable.

If you have any questions or desire further information, please contact me.

Very ruly yours, H. L. Ottoson Manager of Nuclear Operations cc:

D. M. Crutchfield (NRR)

NRC Office of Inspection and Enforcement (Washington, D..C.)