ML20215G704
| ML20215G704 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/12/1987 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| REF-GTECI-A-02, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-OR TAC-8597, NUDOCS 8706230358 | |
| Download: ML20215G704 (2) | |
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'Af Telephone (412) 393-6000 TE*o??*
Shippingport PA 15077.0004 June 12, 1987 U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No..DPR-66' Resolution of USI A-2 " Asymmetric LOCA Loads" Gentlemen:
In our-letter-dated February 19,,1985 we submitted WCAP-10743
" Technical Bases for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for Beaver Valley Unit 1" for NRC review.
- However, at that
- time, we could not justify the leak before break analysis in WCAP-10743 until end of life for BV-1 due to thermal aging concerns.
In our letter dated December 19, 1985 we informed the NRC that Westinghouse had developed alternate' toughness-criteria to address the concerns of thermal aging and had-submitted this methodology as WCAP-10931 " Toughness Criteria for Thermally-Aged Cast Stainless Steel" for NRC review.
In a letter dated December 22, 1986 to Northern States Power Company (Docket Nos. 50-282 and 50-306) the NRC approved the methodology in WCAP-10931 Revision 1
for establishing the fracture criteria for thermally aged' cast stainless i
piping applicable for leak before break analyses.
l In our letter dated June 1, 1987 we submitted proposed Operating License Change Request No.
140 to permit the design of the reactor coolant pump and steam generator supports to be revised in accordance with our submittal.
Included with this submittal was WCAP-11317
" Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural design Basis for Beaver Valley Unit 1" and' a
leakage detection assessment for BV-1.
The leak before break analysis of WCAP-11317 includes the effects of.large bore snubber reduction on the reactor coolant system and uses the methodology of WCAP-10931 Revision 1
for establishing fracture criteria for cast u
~
stainless steel piping at BV-1.
The analysis in WCAP-11317 is l
justified to end of life operation of BV-1.
{
The purpose of this letter is to inform the NRC that WCAP-11317 supersedes WCAP-10743 and along with the leakage detection assessment i
included with our June 1,
1987 submittal, provides the necessary
)
information for NRC resolution of Unresolved-Safety' Issue
.A-2 1
" Asymmetric LOCA Londs" for BV-1.
Therefore,.'we consider our' actions-for USI A-2 comp.ete at this time based on information'provided in our June 1, 1987 submittal.
8706230350 Q h 34 PDR ADOCK FDR D
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Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Resolution of USI A-2 " Asymmetric LOCA Loads" Page 2 If there are any questions concerning this matter, please contact my office.
Very truly yours, e
. D. Sieber Vice President, Nuclear cc:
Mr. F. I. Young, Resident Inspector U.
S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Commission Regional Administrator Region 1 631 Park Avenue King of Prussia, PA 19406 Mr. Peter S. Tam U.
S. Nuclear Regulatory Commission Project Directorate No. 2 Division of PWR Licensing - A Washington, DC 20555
- Mail Stop 316 Addressee only Director, Safety Evaluation & Control Virginia Electric & Power Company P.O. Box 26666 One James River Plaza Richmond, VA 23261