ML18143A955
| ML18143A955 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/07/1976 |
| From: | White L Rochester Gas & Electric Corp |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18143A955 (6) | |
Text
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'iteaoietory Docket Fiie ROCHESTER GAS AND EC TRIC C ION o
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Director of NucleaS~~~hbtSr Regulation ATTN:
Al Schwencer, Chief Operating, Reactors Branch Nl U.S. Nuclear Regulatory Commission Washington, D. C.
20555 go-a+/
Cb Your letter dated June 8, 1976, directed licensees to inform you within 30 days of our schedule for providing an evaluation of the adequacy of the reactor ressure vessel su orts based upon a Request for Additional Information.
That letter was an outcome of your original letter of October,,3.7<<3.975,which informed licen-sees of a generic evaluation oXreactor'pre/s'uj'e vessel supports.
Since October 1975, Rochester Gas and Electric Corporation and several other utilities with operating Westinghouse PWR's have met to jointly investigate several options available to us to address the adequacy of reactor pressure vessel supports in operating plants.
The owners group has considered the performance of analyses similar to those suggested in your Request for Additional Information for each individual plant or for several typical designs which would envelope most of the plants in the owner's group.
Discussions with Westinghouse Electric Corporation have indicated that such analyses would require a significant amount of effort, and in fact would take more than one year but probably less than three years to com-plete.
Plants in the design stage can include those design features which reduce the amount of analysis needed to demonstrate acceptable consequences from postulated events.
Operating plants, on the other
- hand, cannot easily change structural configurations.
Thus, events which were previously analyzed in accordance with approved techniques and found acceptable will now, because of recently developed analysis
- methods, require costly state-of-the-art techniques to demonstrate that the consequences remain within acceptable limits.
We have therefore concluded that our most prudent course of action in response to your questions about our operating plants, including R.
E. Ginna Nuclear Power Plant Unit No. 1, is to propose implementation of an augmented inservice inspection program to preclude a reactor coolant pipe break near the reactor vessel nozzles.
This program will have a positive impact. on plant safety and eliminate the need to perform extensive and lengthy analyses, which have 'minimal impact on the real margin of safety existing in these plants.
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ROCHESTER GAS AND ELECTR ORP.
DATE July 7, 1976 To Al Schwencer
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SHEET NO.
2 Representatives of the owners group and the Westinghouse Elec-tric Corporation met with members of the Nuclear Regulatory Commission Staff on May 25, 1976 to discuss our efforts to that date, including justification for our intention to submit an augmented
- program, and the technical merits of such a program.
Preparation of a report de-tailing the discussion at that meeting is now in progress and should be completed by September 1, 1976.
Upon its completion, that docu-ment will be formally transmitted to you as technical justification for selection of the augmented inservice inspection program.
Sincerely yours, LDW:cern
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