ML111230784
| ML111230784 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/03/2011 |
| From: | Hall J Plant Licensing Branch IV |
| To: | Conklin L Southern California Edison Co |
| Hall, J R, NRR/DORL/LPL4, 301-415-4032 | |
| Shared Package | |
| ML111230775 | List: |
| References | |
| TAC ME5329, TAC ME5330 | |
| Download: ML111230784 (1) | |
Text
REQUEST FOR ADDITIONAL INFORMATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUESTS ISI-3-32, ISI-3-33, AND ISI-3-34 SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 (TAC NOS. ME5329 AND ME5330)
By letter dated January 21, 2011 (Agencywide Documents Access and Management System Accession No. ML110100732), Southern California Edison Company (SCE, the licensee),
submitted a proposed alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI inservice inspection (ISI) requirements regarding examination of certain reactor pressure vessel welds for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. In accordance with 10 CFR 50.55a(a)(3)(i), and the Nuclear Regulatory Commissions (NRCs) safety evaluation approving the use of WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, the NRC staff needs the following information to complete its review of the application.
RAI 1
Regarding the details of the through-wall cracking frequency (TWCF) calculation at SONGS Unit 3, as documented in Table 3 of Proposed Alternative, Enclosure 2, please verify what Chemistry Factor (CF) should be used for the limiting material, region no. 8, intermediate-shell plate, heat no. C-6802-1. If the CF changes for region no. 8, please provide updated outputs in Table 3.
RAI 2
Regarding observed indications from the most recent inservice inspection (ISI) interval examinations at SONGS Unit 3, as documented in Table 2 of Proposed Alternative, Enclosure 2, clearly state the location and size of the one indication that was found in the weld material of the reactor pressure vessel beltline area. Was this indication observed in the 1st and/or 2nd ISI interval inspections? Did the size of the indication change during the course of the subsequent ISI inspections? If there was a change in the size of the indication, can that change be attributed to improved inspection procedures?