ML031400577
| ML031400577 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/09/2003 |
| From: | Salamon G Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LRN-03-0212 | |
| Download: ML031400577 (2) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC MAY O 9 2003 LRN-03-0212 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 SUPPLEMENTAL INFORMATION TO REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION RELIEF REQUEST HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 On February 20, 2003, PSEG Nuclear requested relief (HC-RR-A06) from the requirements of sub paragraph IWA-5250(a)(2) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xi, 1989 Edition, in order to implement the alternative requirements of Code Case N-566-2, Corrective Action for Leakage at Bolted Connections, Section Xl, Division 1, at the Hope Creek Generating Station. The request for relief is for the second 10-year inservice inspection (ISI) interval.
On April 9, 2003, PSEG provided a response to the NRC's request for additional information (RAI) dated April 1, 2003. A telecon was held between the NRC and PSEG on April 28, 2003 to further discuss the information provided in PSEG's response to the NRC's RAI. As a result of this telecon, PSEG is submitting the following supplemental information to support the completion of the NRC Staffs review of relief request HC-RR-A06.
The subject of Question 1 of the NRC's RAI is the use of Code Case N-566-2. If the equipment that is leaking can be isolated, the equipment will be isolated and Code Case N-566-2 would not be invoked. If the leakage cannot be isolated and the system affected causes entry into a Technical Specification Action Statement, Code Case N-566-2 would be invoked and a work plan would be developed for removal and inspection of one bolt at a time. At any time in this process if it is determined that structural integrity cannot be shown to exist, the appropriate actions would be taken in accordance with the plants' Technical Specifications.
The subject of Question 2 of the NRC's RAI is the periodicity of subsequent inspections when Code Case N-566-2 is invoked. If Code Case N-566-2 is used for a leaking 95-2168 REV. 7/99
Document Control Desk 2
MAY 0 9 2003 LRN-03-0212 bolted connection, subsequent inspections of the leaking joint would be scheduled based on the leakage rate from the bolted connection and its affect on the systems or components in the vicinity of the leak, the potential degradation rate of the materials involved, and when the component is scheduled to be out of service for maintenance or testing.
Should you have any questions regarding this submittal please contact Brian Thomas at 856-339-2022.
Sincerely, g~~~
G. Salamon Manager - Nuclear Safety and Licensing C
Mr. H. J. Miller, Regional Administrator U. S. Nuclear Regulatory Commission - Region I 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. G. Wunder, Licensing Project Manager - Hope Creek Mail Stop 08B1 Washington, DC 20555 USNRC Senior Resident Inspector - Hope Creek (X24)
Mr. K. Tosch, Manager, IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625