ML19309E497
| ML19309E497 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 04/17/1980 |
| From: | Vandenburgh D Maine Yankee |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19309E498 | List: |
| References | |
| WMY-80-67, NUDOCS 8004220522 | |
| Download: ML19309E497 (1) | |
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ENGINEERING OFFICE WESTBoRO, MASSACHUSETTS 01581 617-366-9011 B.3.2.1 KMY 80-67 April 17, 1980 United States Nuclear Regulatory Commission Washington, D.C.
20555 Attention: Office of Nuclear Reactor Regulation
References:
- 1) License No. DPR-36 (Docket No. 50-309)
- 2) Maine Yankee letter to USPEC, WMY-80-31, Low Power Steam Line Break, dated February 26, 1980.
- 3) Maine Yankee letter to USNRC, Directorate of Regulatory Operations, Maine Yankee Reportable Occurrence #80-001/01L-0, dated January 30, 1980
Subject:
Revised Steam Line Break Analysis
Dear Sir:
Maine Yankee committed, in Reference 2, to provide a revised steam line break analysis within 30 days following Cycle 5 startup. The revised analysis is required to resolve an incorrect assumption, as identified in Reference 3, in the low power steam line break analysis submitted in Reference 4.
The Attachment provides the required analysis and is intended to replace the steam line break analysis contained in Reference 4 in support of Cycle 5 operation.
In order to resolve the low power steam line break problem identified in Reference 3, the attached analyses were used to determine the required setpoint of the feedwater bypass valve post-trip override to prevent a return to criticality. The revised setpoint was derived based on acceptable post-trip flow rates from the analysis and from a flow test performed during the Cycle 5 startup test program.
In addition, the revised analysis includes the effects of automatic initiation of auxiliary feedwater on low steam generator level as described in Reference 5 as well as the impact of the resulting post-trip override feedwater bypass valve setpoint on the full power steam line break analysis.
8004220
e U.S. Nucl r Regulatory Commission April 17, 1980 Attn: Offica of Nucl r R: actor Regulation P ge 2 Results of the revised analysis indicate that the reactor will remain suberitical following the worst case steam line break for Cycle 5 considering the revised post-trip override feedwater bypass valve setpoint set forth herein.
It is further noted that the reactor will remain subcritical with automatic initiation of the auxiliary feedwater system as described in Reference 5.
We trust that you will find this submittal satisfactory; however, should you desire additional information, feel free to contact us.
Very truly yours, MAIE YANKEE ATOMIC POWER COMPANY DE.Qs D. E. Vandenburgh Vice President RHG/kaf Attachment W
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