ML18153A800
| ML18153A800 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/20/1995 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 94-746A, NUDOCS 9504250121 | |
| Download: ML18153A800 (3) | |
Text
e e
VIRGINIA. ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 20, 1995 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 Serial No.
NL&P/CGL Docket Nos.
License Nos.
CORE UPRATE - MECHANICAL ENGINEERING BRANCH REQUEST FOR ADDITIONAL INFORMATION 94-746A RO 50-280 50-281 DPR-32 DPR-37 The Surry Core Uprate Technical Specification change request was submitted for NRC review by an August 30, 1994 letter (Serial No.94-509).
By a December 16, 1994 letter, the NRC (NRR Mechanical Engineering Branch (EMEB))
requested additional information to support their review of our core uprate submittal.
Our responses to those questions were provided by a February 13, 1995 letter (Serial No.94-746). As a follow-up to that letter, we received an additional question on March 27, 1995. The response to that question is provided by this letter.
If you have further questions or require additional information, please contact us.
Very truly yours,
~Y.01{-~
James P. O'Hanlon Senior Vice President - Nuclear - Response to Request for Additional Information - Proposed Amendment for Power Uprate - Surry Power Station Units 1 and 2 - Mechanical Engineering Branch
'O. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, Virginia 23219
QUESTION:
ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
PROPOSED AMENDMENT FOR POWER UPRATE -
SURRY POWER STATION UNITS 1 AND 2 Mechanical Engineering Branch In Appendix G of the May 2, 1968 Safety Evaluation, on page six, a statement is made that the criteria presented for the design of the reactor internals is satisfactory.
It is assumed that the statement refers to the discussion presented in Section 10 of Supplement - Volume 2.
Is this criteria still valid?
RESPONSE
The information referenced in Section 10 of Supplement - Volume 2 was the response to NRC Question 10.1 on the Surry PSAR. The response to that question identified design criteria, including stress limits, for the original reactor internals design.
The NSSS vendor, Westinghouse, has verified that the information provided in the response to PSAR Question 10.1 with respect to the design criteria, including stress limits, continues to be applicable to the Surry reactor internals. Neither the Surry Improved Fuel design changes nor the Surry Core Uprate Project have a significant impact on the reactor internals. Therefore, the design criteria, including stress limits, applicable to the original design of the internals remain valid at uprated conditions.
1-1