ML112730074

Relief Requests ISI-3-32, ISI-3-33, and ISI-3-34, Alternative to Requirements for Examinations of Welds and Core Support Structure Surfaces, Third 10-Year Inservice Inspection

ML112730074
Person / Time
Site:	San Onofre
Issue date:	11/04/2011
From:	Markley M Plant Licensing Branch IV
To:	Peter Dietrich Southern California Edison Co
Hall, J R, NRR/DORL/LPL4, 301-415-4032
References
TAC ME5329, TAC ME5330
Download: ML112730074 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 4. 2011 Mr. Peter T. Dietrich Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SUBJECT:

SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 - RELIEF REQUESTS ISI-3-32, ISI-3-33, AND ISI-3-34, RELIEF FROM THIRD 10-YEAR INTERVAL INSERVICE INSPECTION FOR CATEGORY B-A, B-D, B-N-2, AND B-N-3 WELDS (TAC NOS. MES329 AND MES330)

Dear Mr. Dietrich:

By letter dated January 7, 2011, as supplemented by letter dated June 9, 2011, Southern California Edison Company (the licensee) submitted Relief Requests (RRs) ISI-3-32, ISI-3-33, and ISI-3-34, requesting relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the third 10-year inservice inspection (lSI) interval for the San Onofre Nuclear Generating Station (SONGS),

Units 2 and 3. In RRs ISI-3-32 and ISI-3-33, the licensee requested, pursuant to paragraph SO.SSa(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR), U.S. Nuclear Regulatory Commission (NRC) approval of proposed alternatives to the inspection requirements for the examination of Category B-A and B-D reactor vessel pressure retaining welds identified in Table IWB-2S00-1 of the ASME Code,Section XI. In RR ISI-3-34, the licensee requested, pursuant to 10 CFR SO.SS(a)(3)(ii), NRC approval of proposed alternatives from the inspection requirements for the examination of accessible interior attachment welds within and beyond the beltline region (Category B-N-2 components) and for a visual examination of the accessible core support structures (Category B-N-3 components) identified in Table IWB-2S00-1 of the ASME Code,Section XI.

Based on the enclosed safety evaluation, the NRC staff concludes that for RRs ISI-3-32 and ISI-3-33, increasing the lSI interval for Category B-A and B-D components from 10 to 20 years will result in no appreciable increase in risk and the proposed alternatives provide an acceptable level of quality and safety. For RR ISI-3-34, the NRC staff concludes that extending the interval for performing the lSI of Category B-N-2 and B-N-3 components from 10 to 20 years is acceptable, as performing these inspections every 10 years as required by Section XI of the ASME Code would result in hardship without a compensating increase in the level of quality and safety. Accordingly, the NRC staff has determined that the licensee has adequately addressed all of the relevant regulatory requirements and that the proposed alternatives provide reasonable assurance of the integrity of the subject welds. Therefore, pursuant to 10 CFR SO.SSa(a)(3)(i) and Oi), the NRC staff authorizes the use of the alternatives described in RRs ISI-3-32, ISI-3-33, and ISI-3-34 at SONGS, Units 2 and 3. The alternatives are authorized until the current operating licenses expire; February 16, 2022, for Unit 2, and November 1S, 2022, for Unit 3.

P. Dietrich - 2 All other ASME Code,Section XI requirements for which alternatives were not specifically requested and approved in these relief requests remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact Mr. Randy Hall at (301) 415-4032 or via e-mail at randy.hall@nrc.gov.

Sincerely.

Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362

Enclosure:

As stated cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS ISI-3-32. ISI-3-33, AND ISI-3-34 SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2AND 3 DOCKET NOS. 50-361 AND 50-362

1.0 INTRODUCTION

By letter dated January 7, 2011 (Reference 1), as supplemented by letter dated June 9, 2011 (Reference 2), Southern California Edison Company (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Paragraph IWB-2412, Inspection Program B, and Table IWB 2500-1, for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. In Relief Requests (RRs) ISI-3-32 and ISI-3-33, the licensee requested U.S. Nuclear Regulatory Commission (NRC) approval of proposed alternatives for SONGS, Units 2 and 3, respectively, pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR),

to extend the inservice inspection (lSI) interval for volumetric examinations of the reactor pressure vessel (RPV) Category B-A and B-D welds from 10 years to 20 years. In RR ISI-3-34, the licensee requested NRC approval, pursuant to 10 CFR 50.55a(a)(3){ii), to perform the visual inspections of B-N-2 and B-N-3 welds (welded core support structures and interior attachment welds) for SONGS, Units 2 and 3, at the same 20-year lSI interval as requested for Category B-A and B-D welds.

2.0 REGULATORY EVALUATION

The lSI of ASME Code Class 1,2, and 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and the applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). The regulations in 10 CFR 50.55a(a)(3) state that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

In accordance with 10 CFR 50.55a(g)(4), the licensee is required to perform inservice examinations of ASME Code Class 1, 2, and 3 components and system pressure tests during the initial10-year interval and subsequent 10-year intervals. These examinations must comply Enclosure

-2 with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(b), subject to the conditions therein.

The current 10-year lSI interval for ASME Code Class 1, 2, and 3 components at SONGS, Units 2 and 3, began on August 18, 2003, and is scheduled to end on August 17, 2013. The Code of record for this inspection interval is the 1995 Edition, including Addenda through 1996, of the ASME Code,Section XI.

2.1 Background The lSI of Category B-A and B-D components consists of visual and ultrasonic examinations intended to determine whether flaws have initiated, pre-existing flaws have extended, or pre existing flaws may have been missed in prior examinations. These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code.

2.2 Summary ofWCAP-16168-NP, Revision 2 In 2006, the Pressurized Water Reactor (PWR) Owners Group submitted topical report WCAP-16168-NP, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 3, referred to as the WCAP in the rest of this document), to the NRC in support of making a risk-informed assessment of extensions to the lSI intervals for Category B-A and B-D components. In the report, the PWR Owners Group justified the proposed extension of the lSI interval for Category B-A and B-D components from 10 to 20 years using data associated with three different PWR plants (referred to as the pilot plants) designed by Westinghouse, Combustion Engineering (CE), and Babcock and Wilcox (B&W).

The WCAP analysis used probabilistic fracture mechanics tools and inputs from the work described in the NRC's pressurized thermal shock (PTS) risk re-evaluation (Reference 4 and 5).

The PWR Owners Group analyses also incorporated the effects of fatigue crack growth and lSI.

Design basis transient data was used as input to the fatigue crack growth evaluation. The effects of lSI were modeled consistent with the previously-approved probabilistic fracture mechanics codes (Reference 6). These effects were accounted for in evaluations performed with the Fracture Analysis of Vessels - Oak Ridge (FAVOR) computer code (Reference 7). All other inputs were identical to those used in the PTS risk re-evaluation.

Based on the results of these studies, the PWR Owners Group concluded that the ASME Code,Section XI, 10-year lSI interval for Category B-A and B-D components in PWR RPVs could be extended to a 20-year inspection interval for any PWR designed by Westinghouse, CE, or B&W as long as the critical, plant-specific parameters described in Appendix A of the WCAP are bound by the parameters of the pilot plants.

2.3 Summary of NRC Safety Evaluation for WCAP The NRC staff concluded in the WCAP safety evaluation (SE) dated July 26, 2011 (Reference 8), that requests to implement alternatives to ASME Code PWR inspection requirements may reference the WCAP methodology, in concert with the guidance provided in Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002

- 3 (Reference 9), and in accordance with the limitations and conditions of the SE. In addition to demonstrating that the critical plant parameters in Appendix A of the WCAP are bounding for a specific plant, licensees must address the following plant-specific information in their request for authorization of an alternative:

1. The dates identified in the request for alternative should be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in the PWR Owners Group letter dated July 10, 2010 (Reference 10). Any deviations from the implementation plan should be discussed in detail in the request for alternative. The proposed alternative lSI interval should be for a maximum of 20 years.

2. The request for alternative lSI interval can use any NRC-approved method to calculate AT30 and RTMAX-X (Reference 5). However, if the request uses the NUREG-1874 methodology to calculate AT 3o , then the request should include the analysis described in paragraph (6) of subsection (f) to the voluntary PTS rule. The analysis should be done for all of the materials in the beltline area with at least three surveillance data points.

3. If the subject plant is a 8&W designed plant, licensees must verify that the fatigue crack growth of 12 heat-up/cool-down transients per year bounds the fatigue crack growth for all of its design basis transients and identify the design basis transients that contribute to significant fatigue crack growth.

4. If the subject plant has RPV forgings that are susceptible to underclad cracking or if the RPV includes forgings with RTMAX-FO (Reference 11) values exceeding 240 degrees Fahrenheit, then the WCAP analyses are not applicable. The licensee must submit a plant-specific evaluation for any extension to the 10-year inspection interval for ASME Code,Section XI, Category 8-A and 8-D RPV welds.

3.0 TECHNICAL EVALUATION

3.1 Description of Proposed Alternatives In RRs ISI-3-32 and ISI-3-33, the licensee proposed to defer the ASME Code-required lSI for Category 8-A and 8-D welds for SONGS, Units 2 and 3, until 2022 (approximately 20 years from the last inspection). This is consistent with the schedule proposed in PWR Owners Group letter dated July 26, 2010 (Reference 10).

For RR ISI-3-34, the licensee proposed that the visual examinations for Category 8-N-2 and 8-N-3 welds be performed consistent with the proposed inspection interval for the Category 8-A and 8-D volumetric examinations.

- 4 3.2 Components for Which Relief is Requested The affected components are the RPV, its interior attachments, and core support structure. The following examination categories and item numbers from IW8-2500 and Table IW8-2500-1 of the ASME Code,Section XI, are addressed in these requests:

For RRs ISI-3-32 and ISI-3-33:

Exam Category Item Number Description 8-A 81.11 Circumferential Shell Welds 8-A 81.12 Longitudinal Shell Welds 8-A 81.21 Circumferential Head Welds 8-A 81.22 Meridional Shell Welds 8-A 81.30 Shell-to-Flange Weld 8-A 81.40 Head-to-Flange Weld 8-0 83.90 Nozzle-to-Vessel Welds 8-0 83.100 Nozzle Inside Radius Section For RR ISI-3-34:

Exam Category Item Number Description 8-N-2 813.50 Interior Attachments within 8eltline Region 8-N-2 813.60 Interior Attachments 8eyond 8eltline Region 8-N-3 813.70 Core Support Structure 3.3 8asis for Proposed Alternatives 3.3.1 RRs ISI-3-32 and ISI-3-33 The basis for ISI-3-32 and ISI-3-33 is found in the NRC-approved version of the WCAP (Reference 12, referred to as WCAP-A). Plant-specific parameters for the subject plant are summarized in Enclosures 1 and 2 to the licensee's letter of January 7, 2011 (Reference 1).

The format of the information is modeled after that found in Appendix A of the WCAP-A.

The licensee concluded that all of the critical parameters for SONGS, Units 2 and 3, listed in Tables 1,2, and 3 of Enclosures 1 and 2 of Reference 1 are bounded by those for the WCAP-A CE pilot plant (Palisades Nuclear Plant) and, therefore, no additional evaluation is required.

3.3.2 RR ISI-3-34 The basis for ISI-3-34 is that performing the visual inspections of the Category 8-N-2 and 8-N-3 components on a different schedule than the examinations for the Category 8-A and 8-0 components would result in significant hardship without a compensating increase in safety. The licensee points out that the Category 8-N-2 and 8-N-3 components have been inspected regularly in the past and no significant indications were noted. Likewise, no significant

-S indications have been identified in the same examinations performed at similar PWRs.

Furthermore, the licensee notes that Category B-N-1 visual inspections and Category B-P pressure tests are performed during each refueling outage and are not affected by this alternative. Finally, the NRC has granted similar relief requests for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (References 12 and 13).

3.4 Duration of Proposed Alternatives The applicability of these alternatives is for the third (currently scheduled to end in 2013) and fourth (currently scheduled to end in 2023) 10-year lSI intervals. The next ASME Category B-A and B-D RPV weld inspections and Category B-N-2 and B-N-3 visual inspections are scheduled for 2022 at SONGS, Units 2 and 3.

3.S NRC Staff Evaluation 3.S.1 RRs ISI-3-32 and ISI-3-33 (SONGS. Units 2 and 3, respectively)

The NRC staff has reviewed Enclosures 1 and 2 to the licensee's letter dated January 7,2011.

The parameters listed in Table 1 of each enclosure were found to be bounded by the WCAP-A.

Furthermore, the RPVs of SONGS, Units 2 and 3, have a single-layer cladding, as assumed in the WCAP-A analysis.

Additional information pertaining to previous and future RPV inspections at SONGS is included in Table 2 of each enclosure. The next volumetric examinations of those components subject to these relief requests for SONGS, Units 2 and 3, are scheduled for 2022; plus or minus one refueling outage. The NRC staff has reviewed the revised PWR Owners Group plan and agrees that the proposed alternatives match the approved inspection plan for the PWR fleet; therefore, the licensee's proposed schedule is acceptable.

With regard to the results of past inspections, there were three indications detected in the most recent (second) interval inspection at Unit 2; all indications were outside the beltline region of the vessel. The licensee evaluated the indications according to ASME Code,Section XI, Paragraph IWB-3S00 and found them to be acceptable without any further evaluation.

Therefore, the NRC staff concludes that the additional information provided in Table 2 of the proposed alternative ISI-3-32 is bounded by the WCAP.

A total of four indications were detected in the most recent (second) interval inspection for SONGS, Unit 3. Only one of these indications was in the high-fluence beltline region. The one indication in the beltline weld metal was located inside of the inner 1 inch, or 1/1 Oth of the RPV thickness. The indication was acceptable according to the requirements of the alternate PTS Rule at 10 CFR SO. 61 a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events"; therefore, there is no requirement for remedial action or further analysis. The remaining three indications were evaluated and found to be acceptable per IWB-3S00 of the ASME Code,Section XI.

Bye-mail dated May 3,2011 (Reference 14), the NRC staff submitted a request for additional information (RAI) regarding the one indication that was found in the beltline weld region. The licensee's RAI response dated June 9, 2011 (Reference 2), included the dimensions of the one

-6 weld metal indication found in the beltline region during the second interval inspection at Unit 3.

The licensee stated that the indication was not found in the first interval inspection. The identification of the beltline indication during the second interval inspection can be attributed to the increased sensitivity of the second inspection. The increased sensitivity is a result of certain requirements of ASME Code,Section XI, Appendix VIII, which were not in effect during the first interval inspection.

8ased on the above, the NRC staff concludes that the beltline region indication found during the Unit 3 second interval inspection may have been a fabrication-induced defect not detected during preservice or previous inservice examinations. There is no evidence that the beltline region indication is growing due to an active aging mechanism; the size of the defect is acceptable per IW8-3500 of the ASME Code,Section XI, and the flaw limits specified in the alternate PTS Rule. Therefore, the NRC staff concludes that the additional information provided in Table 2 of proposed alternative ISI-3-33 is bounded by the WCAP.

The through-wall cracking frequency (TWCF) calculation (Reference 11) was performed using inputs from Table 3 of Enclosures 1 and 2 of the licensee's submittal. The licensee used NRC Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

Position 1.1 methodology to calculate ~T3o in the requests (Reference 15). The NRC staff independently verified the results of the licensee's calculations. The differences between the licensee's and the staff's calculations were insignificant with one exception.

In its e-mail dated May 3,2011 (Reference 14), the NRC staff requested additional information regarding the chemistry factor for the SONGS, Unit 3 Region 8 intermediate shell plate, heat no. C-6802-1. The licensee's RAI response dated June 9,2011 (Reference 2), described how the chemistry factor for Region 8 was calculated from credible surveillance data per RG 1.99, Revision 2, Position 2.1 and not Position 1.1 as indicated in the initial relief request. A revised version of Table 3 from Enclosure 2 was included to reflect this corrected information.

The NRC staff concludes that the details of the TWCF calculations for the SONGS, Units 2 and 3, proposed alternatives ISI-3-32 and ISI-3-33 are bounded by the WCAP and, therefore, the proposed alternatives are acceptable.

3.5.2 RR ISI-3-34, SONGS, Units 2 and 3 The NRC staff has reviewed RR ISI-3-34 for SONGS, Units 2 and 3 (Enclosure 3 of Reference 1), and agrees that this request is identical to the requests described in References 12 and 13 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The Category 8-N-2 and 8-N-3 inspections can only be performed after the removal of fuel and internals from the RPV, usually at the same time as the lSI of Category 8-A and 8-D components. None of the previous inspections of Category 8-N-2 and 8-N-3 components at SONGS, Units 2 and 3, have identified significant indications of cracking or other problems. No evidence of service-induced degradation has been noted in the same inspections at other similar nuclear power plants. The staff notes that no other nuclear power plant performs the lSI for Category 8-N-2 and 8-N-3 components on a different interval schedule than that for the lSI for Category 8-A and 8-D welds. Furthermore, the NRC staff agrees that the change in interval for the inspections does not impact the defense-in-depth philosophy. Hence, the alternative proposed is acceptable under the provisions of 10 CFR 50.55a(a)(3)(ii).

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4.0 CONCLUSION

The NRC staff has completed its review of RRs ISI-3-32, ISI-3-33, and ISI-3-34 for SONGS, Units 2 and 3. The staff concludes that increasing the lSI interval for Category B-A and B-D components from 10 to 20 years will result in no appreciable increase in risk. This conclusion is based on the fact that the plant-specific information provided by the licensee is bounded by the data in the WCAP-A, and the requests meet all of the conditions and limitations described in the WCAP-A On that basis, the staff concludes that RRs ISI-3-32 and ISI-3-33 provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the use of RRs ISI-3-32 and ISI-3-33 for SONGS, Units 2 and 3, respectively, for Category B-A and B-D components. This authorization will remain in effect until the current operating licenses expire; February 16, 2022, for SONGS, Unit 2, and November 15, 2022, for Unit 3.

For RR ISI-3-34, the NRC staff concludes that performing the lSI of Category B-N-2 and B-N-3 components every 10 years as required by Section XI of the ASME Code would result in hardship without a compensating increase in the level of quality and safety. The staff agrees that the Category B-N-2 and B-N-3 inspections should be performed at the same time as the Category B-A and B-D component inspections, as proposed. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the use of RR ISI-3-34 for SONGS, Units 2 and 3, until the current operating licenses expire; February 16, 2022, for SONGS, Unit 2, and November 15, 2022, for Unit 3.

All other requirements of the ASME Code,Section XI, for which alternatives were not specifically requested and approved in these relief requests remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

5.0 REFERENCES

1. St. Onge, R. J., Southern California Edison, letter to U.S. Nuclear Regulatory Commission, 'Third Ten-Year Inservice Inspection (lSI) Interval, 10CFR50.55a Requests ISI-3-32, ISI-3-33, and ISI-3-34, San Onofre Nuclear Generating Station, Units 2 and 3,"

dated January 7,2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110100732).

2. St. Onge, R. J., Southern California Edison, letter to U.S. Nuclear Regulatory Commission, "Third Ten-Year Inservice Inspection (lSI) Interval, 10CFR50.55a Requests ISI-3-32, ISI-3-33, and ISI-3-34, San Onofre Nuclear Generating Station, Units 2 and 3,"

dated June 9,2011 (ADAMS Accession No. ML11161A156).

3. PWR Owners Group, WCAP-16168-NP, Revision 2, "(DRAFT) Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," October 2007 (ADAMS Accession No. ML072920413).

4. U.S. Nuclear Regulatory Commission, 'Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report," NUREG-1806, dated May 24,2006 (ADAMS Accession No. ML061580318).

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5. U.S. Nuclear Regulatory Commission, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," NUREG-1874, dated March 1,2007 (ADAMS Accession No. ML070860156).

6. Westinghouse Electric Company, WCAP-14572, Revision 1-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," February 1999 (ADAMS Accession No. ML012630375).

7. Oak Ridge National Laboratory and U.S. Nuclear Regulatory Commission, ONRLlNRC/LTR-04/18, "Electronic Archival of the Results of Pressurized Thermal Shock Analyses for Beaver Valley, Oconee, and Palisades Reactor Pressure Vessels Generated with the 04.1 Version of FAVOR," dated October 15,2004 (ADAMS Accession No. ML042960391).

8. Nelson, R. A., U.S. Nuclear Regulatory Commission, letter to W. Anthony Nowinowski, PWR Owners Group, "Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP-A, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval'," dated July 26,2011 (ADAMS Accession No. ML111610242).

9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis," November 2002 (ADAMS Accession No. ML023240437).

10. Arey, Jr., M. L., PWR Owners Group, letter to U.S. Nuclear Regulatory Commission, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,' (PA-MSC-0120)," dated July 12, 2010, (ADAMS Accession No. ML11153A033).

11. Buschbaum, D. E., PWR Owners Group, letter to U.S. Nuclear Regulatory Commission, "Transmittal of NRC Approved Topical Report WCAP-16168-NP-A, Rev. 2, 'Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval,' (TAC No. MC9768) (MUHP 5097/5098/5099, Task 2008/2059, PA MSC-0120)," dated June 13, 2008 (ADAMS Accession No. ML082820046).

12. Flaherty, M. D., Constellation Energy, letter to U.S. Nuclear Regulatory Commission, "Calvert Cliffs Nuclear Power Plant, Unit No.2; Docket No. 50-318, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-020 and ISI-021 )," dated October 1, 2008 (ADAMS Accession No. ML082760280).

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13. Flaherty, M. D., Constellation Energy, letter to U.S. Nuclear Regulatory Commission, "Calvert Cliffs Nuclear Power Plant, Unit No.1; Docket No. 50-317, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)," dated February 18, 2009 (ADAMS Accession No. ML090540062).

14. Hall, J. R., U.S. Nuclear Regulatory Commission, electronic mail to Linda Conklin, Southern California Edison, "Draft Request for Additional Information - Relief Requests ISI-3-32 through ISI-3-34 for San Onofre Nuclear Generating Station (TAC Nos. ME5329 and ME5330)", dated May 3,2011 (ADAMS Accession No. ML111230775).

15. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (ADAMS Accession No. ML003740284).

Principal Contributor: P. Purtscher, NRR/DCI Date: November 4, 2011

P. Dietrich -2 All other ASME Code,Section XI requirements for which alternatives were not specifically requested and approved in these relief requests remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact Mr. Randy Hall at (301) 415-4032 or via e-mail at randy.hall@nrc.gov.

Sincerely, IRAI Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362

Enclosure:

As stated cc w/encl Distribution via Listserv DISTRIBUTION:

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