ML19093B042
| ML19093B042 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/07/1978 |
| From: | Stallings C Virginia Electric & Power Co (VEPCO) |
| To: | Case E, Stello V Office of Nuclear Reactor Regulation |
| References | |
| Serial No. 067/012578 | |
| Download: ML19093B042 (1) | |
Text
e March. 7, 1978
- Mr. Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Victor Stello, Jr. Director Division of Operating Reactors
Dear Mr. Case:
License Nos. DPR-,32 DPR..;37' This is in response to your letter of January 25, 1978, concerning the design of the reactor vessel support system for Surry Power Station Unit Nos.
1 and 2.
Your letter requested a response within 30 days indicating our in-tent to proceed with an evaluation of the overall asymmetric loss of coolant accident loads as described in your letter and attachments.
You also request-ed that within 90 days we submit our detailed schedule for providing the re-quired evaluation.
This.letter is to notify you that a task group of utilities with West-inghouse plants has been formed, of which Virginia Electric and Power Com-pany is a participant, to examine the complexity of primary system main cool-ant piping system breaks as outlined in your January 25, 1978 letter and to identify similarities between Westinghouse plants for the purpose of deter-mining a consistent.evaluation of this issue.
The task group has met with Westinghouse and with the NRC during the past several months for the purpose of developing a program that is acceptable to both parties. Preliminary work is proceeding which will allow the task group to outline a realistic program within the 90 days you requested.
The exact content of this program is not known at this time, but its purpose will be to re-assure the *original plant des.ign and assure the safety of the plant should a pipe break occur at specified locations in the reactor coolant system.
The particular resolution of this issue will vary from plant to plant, but as a minimum it will include an analytical evaluation that assesses the safety of the plant. If needed, this evaluation may be supple-mented by a plant modification, probability analysis or augmented inservice inspection.
cc:
Mr. James P. O'Reilly
.Very truly yours,
'Z/J, }J?. ?JnC?=---~~r.::v C. M. Stall~ngs Vice President-Power Supply and Production Operations