ML20235Z639
| ML20235Z639 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/10/1989 |
| From: | Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20235Z641 | List: |
| References | |
| TASK-15-19, TASK-RR 5000-89-1733, NUDOCS 8903160045 | |
| Download: ML20235Z639 (2) | |
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GPU Nuclear Corporation UC GSr One Upper Pond Road Parsippany, New Jersey 07054 201-316-7000 TELEX 136-482 Writer's Direct Dial Number:
j March 10, 1989 5000-89-1733 i
U. S. Nuclear Regulatory Commission Attention:
Document Control Desk i
Mail Station P1-137 Washington, DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station (OCNGS) i Docket No. 50-219 Main Steam Isolation Valves (MSIV)
GPU Nuclear met with the staff on February 13,19B9 to resolve outstanding concerns related to the restart of the OCNGS from its 12R refueling outage.
The inability of the MSIVs to maintain their specified leakage rates for the time period assumed in Design Basis Accident (DBA) scenarios was addressed and is the subject of this correspondence.
The identified concern arises when the MSIV Technical Specification leakage requirement is compared to the anticipated performance of the valves / plant in an accident condition. The 10 CFR50, Appendix J Type C MSIV testing is performed with control air pressure applied to the valve actuator. The non-safety grade normal instrument air / nitrogen system would be unavailable during DBA conditions.
As a result valve seating force would be less than that provided during the Type C test.
In order to address the safety significance, GPU Nuclear calculated MSIV leakage as a function of both drywell pressure and valve actuator pressure varying with time. This method is consistent with the staff's recommendation to perform a more realistic analysis factoring in the effects of drywell pressure versus MSIV leakage rates as a function of time. The recommendation was made concerning SEP Topic XV-19 "LOCAs Resulting from spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary".
The attached Topical Report (TR) " Evaluation of Oyster Creek Main Steam Isolation Valve Leak Rate" describes this methodology in detail.
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l The NRC staff also requested that GPUN list the major assumptions used in the analysis. They are: 1. a drop in containment pressure as depicted in FSAR I
Figure 6.2-3, and; 2. that the valves tested, which are located outside the d
drywell, have similar leakage characteristics as the valves located inside l
the drywell. The TR discusses each of these assumptions in detail.
Using the methodology described in the attached TR, integrated leakage /off-site dose associated with zero valve actuator pressure is below the SEP integrated leakage /off-site dose.
Based on the above, anticipated MSIV leakage under DBA conditions does not represent a safety concern.
Very t uly yours, t
NYW R. F. Wi son Director, Technical Functions RFW:crb Enclosure cc:
Regional Administrator Region i U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 i
Resident Inspector i
Oyster Creek Nucioar Generating Station Mr. Alex Dromerick U.S. Nuclear Regulatory Commission Mail Station P1-137 Washington, DC 20555 i
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