ML110100732

Third Ten-Year Inservice Inspection (151) Interval 10CFR50.55a Requests 151-3-32, 151-3-33, and 151-3-34 San Onofre Nuclear Generating Station, Units 2 and 3

ML110100732
Person / Time
Site:	San Onofre
Issue date:	01/07/2011
From:	St-Onge R J Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML110100732 (21)

Text

SOUTH"!N CAUFORNIAEDISONAn E D ISON I N T ER N ATIO NALCompany January7,2011 ATTN: DocumentControlDeskU.S.Nuclear Regulatory Commission Washington

,DC20555-0001 Richard J.St.Onge Dir e ctor Nuclear Regulatory Affairs 10CFR50.55a

Subject:

, DocketNos.50-361and 50-362 Third Ten-Year Inservice Inspection (151)Interval 10CFR50.55a Requests151-3-32,151-3-33, and 151-3-34 San Onofre Nuclear Generating Station,Units2and3DearSirorMadam

, Pursuantto10CFR 50.55a(a)(3)(i)

, Southe rn CaliforniaEdison(SCE)requestsNRC approva lofthefo llowingrequestfortheSanOnofreNuc lear GeneratingStation(SONGS)

Third Ten-Year IntervallnserviceInspection(lSI)Program:Extensionofthe intervalfor volumetric examination of essentially 100%ofreactorvesselpressu re retaining,Examination Category B-A a ndB-Dwelds,from10yearsto20years.App rova lofthisrequestwould result inareductioninman-remexposu re and examinationcosts.Enclosure1providesthedetailsof thisrequestasThird Ten-YearlSI1OCFR50

.55arequestISI 32forUnit2.Enclosure2providesthedetailsofthisrequestasThirdTen-YearlSI1OCFR50.55arequest ISI-3-33forUnit3.'Pursuant to 10CFR50.55a(a)(3)(ii),SCEalsorequestsNRCapprovalofthefollow ingrequestfortheSONGSThird Ten-YearIntervallSIProgram

Extensionoftheinte rval from10to20yearsforv isual exam inationunderASMESectionXITableIWB-2500-1 examination categoriesB-N-2andB-N-3

,ItemNos.B13.50,B13.60 ,andB13.70.Enclosure3ofthis lette r prov idesthede tailsofthisrequestasThird Ten-Yea rlSI1 OCFR50.55a request ISI-334fo rUnits2and3.

P.O.B ox 128SanClemente,CA92674 DocumentControlDesk-2-January 7, 2011 This letterandthe Enclosurescontainnonew commitments.Shouldyouhaveany questions, please contactMs.Linda 1.Conklinat(949) 368-9443.Sincerely, Enclosure1:ISI-3-32forUnit2 : ISI-3-33forUnit3 Enclosure 3: ISI-3-34forUnits2and3 cc: E.E.Collins,Regional Administrator,NRCRegionIVR.Hall,NRC ProjectManager,San OnofreUnits2and3G.G.Warnick, NRC Senior Resident Inspector,SanOnofreUnits2and3

Enclosure 1 10 CFR 50.55a Request ISI-3-32 Proposed Alternative in Accord ance with 10 CFR 50.55a(a)(3)(i)

Proposed Alternative Provides Acceptable Level of Quality or Safety San Onofre Nuclear Generating Station, Unit 2 San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 1 of 6 1. ASME Code Component(s) Affected The affected component is the San Onofre Nuclear Generating Station (SONGS) Unit 2 reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference

1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.

Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Shell Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section

(Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval. The SONGS Unit 2 third 10-year in-service inspection interval ends in 2013. The applicable Code for the fourth 10-year in-service inspection interval will be selected in accordance with the rules of 10 CFR 50.55a.

San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 2 of 6 4. Reason for Request An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval. Extension of time frame between Examination Category B-A and B-D welds from 10 years to 20 years will result in a reduction in man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use Southern California Edison proposes to not perform the ASME Code required volumetric examination of the SONGS Unit 2 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third in-service inspection interval. These examinations would need to be performed in 2012 without approval of this relief request. Southern California Edison will perform the ASME Code required volumetric examination of the SONGS Unit 2 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the fourth in-service inspection interval in 2022. These dates are within one refueling outage of the dates provided in PWR Owners Group letter OG-06-356 (Reference

2) and the revised implementation plan (Reference

3). In accordance with IWA-2430(d)(1), the proposed inspection date of 2022 may be reduced or extended by as much as one year.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current time frame can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 4).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 2, "Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 5). This study focuses on risk assessments of materials within beltline region of the RV wall. The results of the time frame calculations for SONGS Unit 2 were compared to those obtained from the CE pilot plant evaluated in WCAP-16168-NP-A, Revision 2. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for SONGS Unit 2 are bounded by the results of the CE pilot plant qualifies SONGS Unit 2 for an ISI interval extension. Table 1 below lists these critical parameters investigated in the WCAP and compares the results of the CE pilot plant to that SONGS Unit 2. Tables 2 and 3 provide additional information that was requested by NRC and included in Appendix A of Reference

5.

San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 3 of 6 Table 1 : Critical Parameters for Application of Bounding Analysis - SONGS Unit 2 Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required? Dominant Pressurized Thermal Shock (PTS) Transients in the NRC PTS Risk Study are Applicable NRC PTS Risk Study (Reference

6) PTS Generalization Study (Reference

7) No Through-Wall Cracking Frequency (TWCF) 3.16E-7 Events per year (Reference

5) 9.24E-13 Events per year (Calculated per Reference

5) No Frequency and Severity of Design Basis Transients 13 heatup/cooldown cycles per year (Reference

5) Bounded by 13 heatup/cooldown cycles per year No Cladding Layers (Single/Multiple) Single Layer (Reference 5) Single Layer No Table 2 below provides a summary of the latest reactor vessel inspection for SONGS Unit 2 and evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the SONGS Unit 2 reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection - SONGS Unit 2 Inspection methodology: During the most recent in-service inspection, all Category B-A shell-to-shell welds, including those located within the beltline region of the RV, were inspected according to ASME Section XI Appendix VIII 1995 Edition with the 1996 Addenda, as modified by 10CFR50.55a(b)(2)(xiv, xv and xvi) (Reference 1) requirements. All other Category B-A and B-D weld inspections were performed to ASME Section XI and Section V 1989 Edition requirements for examinations applicable to operational Interval 2. USNRC Regulatory Guide 1.150, Revision 1 (Reference 8) was applicable to flange-to-shell and nozzle examinations from the nozzle bore requirements. Future in-service inspections will be performed to ASME Section XI Appendix VIII requirements. Number of past inspections: Two 10-Year inservice inspections have been performed. Number of indications found: There were three recordable indications identified during the most recent in-service inspection but none were in the beltline region. The recorded indications were evaluated per ASME Code Section XI, 1989 Edition IWB-3510 and found to be acceptable with no further evaluation required. Proposed inspection schedule for balance of plant life: The third in-service inspection must be performed in 2012. Pending approval of this relief request, this inspection will be performed in 2022. These dates are within one refueling outage of the dates provided in OG-06-356 (Reference 2) and the revised implementation plan (Reference 3).

San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 4 of 6 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation - SONGS Unit 2 at 54 Effective Full Power Years (EFPY)

Inputs Reactor Coolant System Temperature, T RCS[°F]:N/A Twall [inches]: 8.84375 No. Region/Component Description Material Heat No./Type Cu (1) [wt%] Ni (1) [wt%] R.G. 1.99 Pos. CF (2) [°F] RTNDT(u)(1) [°F] Fluence [10 19 Neutron/cm 2 , E > 1.0 MeV]

1 Inter. Shell Long. Weld 2-203 A E8018 0.03 0.90 1.1 41 -60 7.27 2 Inter. Shell Long. Weld 2-203 B E8018 0.03 0.91 1.1 41 -60 7.27 3 Inter. Shell Long. Weld 2-203 C E8018 0.03 0.95 1.1 41 -60 7.27 4 Lower Shell Long. Weld 3-203 A 83637 0.05 0.12 1.1 40 -50 7.38 5 Lower Shell Long. Weld 3-203 B 83637 0.04 0.06 1.1 30 -50 7.38 6 Lower Shell Long. Weld 3-203 C 83637 0.06 0.11 1.1 42 -50 7.38 7 Lower-Inter. Shell Girth Weld 9-203 90130 0.07 0.29 1.1 69 -60 7.38 8 Intermediate Shell C-6404-1 SA-533B Cl.1 0.10 0.56 1.1 65 20 7.27 9 Intermediate Shell C-6404-2 SA-533B Cl.1 0.10 0.59 1.1 65 20 7.27 10 Intermediate Shell C-6404-3 SA-533B Cl.1 0.10 0.56 1.1 65 20 7.27 11 Lower Shell C-6404-4 SA-533B Cl.1 0.10 0.62 1.1 65 20 7.38 12 Lower Shell C-6404-5 SA-533B Cl.1 0.11 0.64 1.1 75 10 7.38 13 Lower Shell C-6404-6 SA-533B Cl.1 0.10 0.58 1.1 65 -10 7.38 Outputs Methodology Used to Calculate T 30:Regulatory Guide 1.99, Revision 2 (3) Controlling Material Region No. (From Above) RTMAX-XX [°R] Fluence [10 19 Neutron/cm 2 , E > 1.0 MeV]

FF (Fluence Factor) T 30 [°F] TWCF 95-XX Limiting Axial Weld - AW 12 580.0 7.38 1.471 110.34 0.00E+00 Limiting Plate - PL 12 580.0 7.38 1.471 110.34 3.70E-13 Circumferential Weld - CW 12 580.0 7.38 1.471 110.34 0.00E+00 TWCF95-TOTAL (AWTWCF95-AW + PLTWCF 95-PL + CWTWCF95-CW): 9.24E-13 (1) Reference 9 (2) Reference 10 (3) Reference 11 San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 5 of 6 6. Duration of Proposed Alternative This request is applicable to the SONGS Unit 2 inservice inspection program for the third and fourth 10-year inspection intervals.

Precedents:

Letter from M. D. Flaherty , (Calvert Cliffs Nuclear Power Plant) to Document Control Desk (NRC), dated February 18, 2009 (ADAMS Accession No.

ML090540062);

Subject:

"Unit 1; Docket No. 50-317, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)

Letter from N. L. Salgado (NRC) to J. A. Spina (Calvert Cliffs Nuclear Power Plant, dated November 9, 2009 (ADAMS Accession No. ML093030052);

Subject:

"Relief Requests Inservice Inspection (ISI)-022 and ISI-023 to Extend Reactor Vessel and

Reactor Vessel Internal Weld Examinations - Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC Nos. ME0668 and ME0669)

7. References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition with the 1996 Addenda, American Society of Me chanical Engineers, New York. 2. OG-06-356, "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." MUHP 5097-99, Task 2059," October 31, 2006. 3. OG-09-454, "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," December 1, 2009. 4. NRC 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. 5. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 2008. 6. NUREG-1874, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," 10/3/07 (ADAMS Accession Number ML070860156).

San Onofre Unit 2 10 CFR 50.55a Request ISI-3-32 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety 6 of 6 7. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482). 8. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inse rvice Examinations," February 1983. 9. BAW-2408, "Analysis of the 263

° Capsule Southern California Edison Company San Onofre Unit 2 Nucl ear Generating St ation," 10/2001. 10. WCAP-16005-NP, Revision 6, "San Onofre Nuclear Generating Station Unit 2 RCS Pressure and Temperature Limits Report", 10/2008. 11. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.

Enclosure 2 10 CFR 50.55a Request ISI-3-33 Proposed Alternative in Accord ance with 10 CFR 50.55a(a)(3)(i)

Proposed Alternative Provides Acceptable Level of Quality and Safety San Onofre Nuclear Generating Station, Unit 3 San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 1 of 6 1. ASME Code Component(s) Affected The affected component is the San Onofre Nuclear Generating Station (SONGS) Unit 3 reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference

1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.

Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Shell Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section

(Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval. The SONGS Unit 3 third 10-year in-service inspection interval ends in 2013. The applicable Code for the fourth 10-year in-service inspection interval will be selected in accordance with the rules of 10 CFR 50.55a.

San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 2 of 6

4. Reason for Request

An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval. Extension of time frame between Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use Southern California Edison proposes to not perform the ASME Code required volumetric examination of the SONGS Unit 3 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third in-service inspection interval. These examinations wo uld need to be performed in 2012 without approval of this relief request. Southern California Edison will perform the ASME Code required volumetric examination of the SONGS Unit 3 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the fourth in-service inspection interval in 2022. These dates are within one refueling outage of the dates provided in PWR Owners Group letter OG-06-356 (Reference

2) and the revised implementation plan (Reference

3). In accordance with IWA-2430(d)(1), the proposed inspection date of 2022 may be reduced or extended by as much as one year.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current time frame can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 4).

The methodology used to conduct this analysi s is based on that defined in the study WCAP-16168-NP-A, Revision 2, "Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 5). This study focuses on risk assessments of materials within beltline region of the RV wall. The results of the time frame calculations for SONGS Unit 3 were com pared to those obtained from the CE pilot plant evaluated in WCAP-16168-NP-A, Revision 2. Appendix A of the WCAP identifies the parameters to be compared. Demonstratin g that the parameters for SONGS Unit 3 are bounded by the results of t he CE pilot plant qualifies SONGS Unit 3 for an ISI interval extension. Table 1 below lists these critical parameters investigated in the WCAP and compares the results of the CE pilot plant to that SONGS Unit 3. Tables 2 and 3 provide ad ditional information that was requested by NRC and included in Appendix A of Reference

5.

San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 3 of 6 Table 1 : Critical Parameters for Application of Bounding Analysis - SONGS Unit 3 Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required? Dominant Pressurized Thermal Shock (PTS) Transients in the NRC PTS Risk Study are Applicable NRC PTS Risk Study (Reference

6) PTS Generalization Study (Reference

7) No Through-Wall Cracking Frequency (TWCF) 3.16E-7 Events per year (Reference

5) 6.78E-12 Events per year (Calculated per Reference

5) No Frequency and Severity of Design Basis Transients 13 heatup/cooldown cycles per year (Reference

5) Bounded by 13 heatup/cooldown cycles per year No Cladding Layers (Single/Multiple) Single Layer (Reference 5) Single Layer No Table 2 below provides a summary of the latest reactor vessel inspection for SONGS Unit 3 and evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the SONGS Unit 3 reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection - SONGS Unit 3 Inspection methodology: The most recent in-service inspection of the Category B-A and B-D welds was performed to ASME Section XI Appendix VIII 1995 Edition with the 1996 Addenda, as modified by 10CFR50.55a(b)(2)(xiv, xv and xvi) (Reference 1). USNRC Regulatory Guide 1.150, Revision 1 (Reference 8) was applicable to the flange-to-shell weld as examined manually from the flange seal surface. Future in-service inspections will be performed to ASME Section XI Appendix VIII requirements. Number of past inspections: Two 10-Year in-service inspections have been performed. Number of indications found: There were four recordable indications identified during the most recent in-service inspection, but only one indication was in the beltline region. This indication was acceptable per Table IWB-3510 of Section XI of the ASME Code. The indication is within the inner 1/10 th or 1" of the reactor vessel thickness, has a through-wall extent of 0.18", and is located within the weld material of the RV beltline. Twenty three indications of this size would be allowable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 9). The remaining three indications were evaluated per ASME Code IWB-3510 for a Category B-A weld, IWB-3512 for a B-D weld, and IWB-3514 for a B-J weld and found to be acceptable with no further evaluation required. Proposed inspection schedule for balance of plant life: The third in-service inspection must be performed in 2012. Pending approval of this relief request, this inspection will be performed in 2022. These dates are within one refueling outage of the dates provided in OG-06-356 (Reference 2) and the revised implementation plan (Reference 3).

San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 4 of 6 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation - SONGS Unit 3 at 54 Effective Full Power Years (EFPY)

Inputs Reactor Coolant System Temperature, T RCS[°F]:N/A Twall [inches]: 8.84375 No. Region/Component Description Material Heat No./Type Cu (1) [wt%] Ni (1) [wt%] R.G. 1.99 Pos. CF (2) [°F] RTNDT(u)(1) [°F] Fluence [10 19 Neutron/cm 2 , E > 1.0 MeV]

1 Inter. Shell Long. Weld 2-203 A 83650 0.04 0.17 1.1 40 -40 6.73 2 Inter. Shell Long. Weld 2-203 B 83650 0.05 0.21 1.1 50 -40 6.73 3 Inter. Shell Long. Weld 2-203 C 83650 0.04 0.08 1.1 32 -40 6.73 4 Lower Shell Long. Weld 3-203 A 88114 0.04 0.21 1.1 44 -70 6.52 5 Lower Shell Long. Weld 3-203 B 88114 0.04 0.19 1.1 42 -70 6.52 6 Lower Shell Long. Weld 3-203 C 88114 0.04 0.21 1.1 44 -70 6.52 7 Lower-Inter. Shell Girth Weld 9-203 90069 and 90144 0.06 0.04 1.1 34 (3) -50 (3) 6.73 8 Intermediate Shell C-6802-1 SA-533B Cl.1 0.06 0.58 1.1 72 40 6.73 9 Intermediate Shell C-6802-2 SA-533B Cl.1 0.04 0.57 1.1 26 10 6.73 10 Intermediate Shell C-6802-3 SA-533B Cl.1 0.06 0.58 1.1 37 20 6.73 11 Lower Shell C-6802-4 SA-533B Cl.1 0.05 0.56 1.1 31 10 6.52 12 Lower Shell C-6802-5 SA-533B Cl.1 0.04 0.55 1.1 26 10 6.52 13 Lower Shell C-6802-6 SA-533B Cl.1 0.06 0.62 1.1 37 20 6.52 Outputs Methodology Used to Calculate T 30:Regulatory Guide 1.99, Revision 2 (4) Controlling Material Region No. (From Above) RTMAX-XX [°R] Fluence [10 19 Neutron/cm 2 , E > 1.0 MeV]

FF (Fluence Factor) T 30 [°F] TWCF 95-XX Limiting Axial Weld - AW 8 604.5 6.73 1.457 104.87 0.00E+00 Limiting Plate - PL 8 604.5 6.73 1.457 104.87 2.71E-12 Circumferential Weld - CW 8 604.5 6.73 1.457 104.87 0.00E+00 TWCF95-TOTAL (AWTWCF95-AW + PLTWCF 95-PL + CWTWCF95-CW): 6.78E-12 (1) Reference 10 (2) Reference 11 (3) Maximum value for two heats of material (4) Reference 12 San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 5 of 6

6. Duration of Proposed Alternative This request is applicable to the SONGS Unit 3 inservice inspection program for the third and fourth 10-year inspection intervals.

Precedents:

Letter from M. D. Flaherty , (Calvert Cliffs Nuclear Power Plant) to Document Control Desk (NRC), dated February 18, 2009 (ADAMS Accession No.

ML090540062);

Subject:

"Unit 1; Docket No. 50-317, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)

Letter from N. L. Salgado (NRC) to J. A. Spina (Calvert Cliffs Nuclear Power Plant, dated November 9, 2009 (ADAMS Acce ssion No. ML093030052);

Subject:

"Relief Requests Inservice Inspection (ISI)-022 and ISI-023 to Extend Reactor Vessel and

Reactor Vessel Internal Weld Examinations - Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC Nos. ME0668 and ME0669)

7. References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition with the 1996 Addenda, American Society of Me chanical Engineers, New York. 2. OG-06-356, "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." MUHP 5097-99, Task 2059," October 31, 2006. 3. OG-09-454, "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," December 1, 2009. 4. NRC 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. 5. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 2008. 6. NUREG-1874, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," 10/3/07 (ADAMS Accession Number ML070860156).

San Onofre Unit 3 10 CFR 50.55a Request ISI-3-33 In Accordance with 10 CFR 50.55a(a)(3)(i) Proposed Alternative Provides Acceptable Level of Quality and Safety Page 6 of 6 7. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482). 8. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inse rvice Examinations," February 1983. 9. Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. 1, dated January 4, 2010 and No. 22 with corrections to part (g) dated February 3, 2010. 10. BAW-2454, "Analysis of the 263

° Capsule Southern California Edison Company San Onofre Unit 3 Nucl ear Generating St ation," 1/2004. 11. WCAP-16167-NP, Revision 2, "San Onofre Nuclear Generating Station Unit 3 RCS Pressure and Temperatur e Limits Report", 10/2008.

12. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.

Enclosure 3 10 CFR 50.55a Request ISI-3-34 Proposed Alternative in Accord ance with 10 CFR 50.55a(a)(3)(ii) Hardship without a Compensating Increase in the Level of Quality and Safety San Onofre Nuclear Generating Station, Units 2 and 3

San Onofre Units 2 and 3 10 CFR 50.55a Request ISI-3-34 In Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship without Compensating Increase in the Level of Quality and Safety Page 1 of 4 American Society of Mechanical Engineers (ASME) Code Component(s)

Affected The affected components are the San Onofre Nuclear G enerating Station (SONGS)

Units 2 and 3 reactor pressure vessels (RPV s), specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the RPVs. These examination categories and item numbers are from IWB-2500 and Tabl e IWB-2500-1 of the ASME BPV, Code Section XI.

Examination Category Item No. Description B-N-2 B13.50 Interior Attachm ents Within Beltline Region B-N-2 B13.60 Interior Attach ments Beyond Beltline Region B-N-3 B13.70 Core Support Structure (Throughout this request the above examinat ion categories are referred to as "the subject examinations" and the ASME BPV Code Section XI is referred to as "the Code".)

Applicable Code Edition and Addenda

ASME Code Section XI, "Rules for Inserv ice Inspection of Nuclear Power Plant Components," Code 1995 Edition with the 1996 Addenda.

Applicable Code Requirement

In accordance with IWA-2430(d)(1), each inspection interval may be reduced or extended by as much as one year. Adjust ments shall not cause successive intervals to be altered more than one year from the original pattern of intervals.

Additionally, Table IWB-2500-1, Examinat ion Categories B-N-2 and B-N-3, Item Numbers B13.50, B13.60, and B 13.70 requires a visual exam ination of the accessible interior attachment welds within a nd beyond the beltline region and a visual examination of the ac cessible core support structure su rfaces of the RPV once each ten-year interval. The curr ent SONGS Unit 2 and Unit 3 third 10-year ISI interval began on August 18, 2003 and is scheduled to end on August 17, 2013.

Reason for Request

In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the Pressurized Water Reactor Owners Group prov ided the technical and regulatory basis for decreasing the frequency of inspections by extending the ASM E Code Section XI ISI interval from the current 10 years to 20 years for ASME Code Section XI Examination Categories B-A and B-D RPV welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 200 8 (Reference 3).

San Onofre Units 2 and 3 10 CFR 50.55a Request ISI-3-34 In Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship without Compensating Increase in the Level of Quality and Safety Page 2 of 4 To implement the change presented in Refer ence 2, we are submitting Enclosures 1 and 2 (ISI-3-32 and ISI-3-33 for Units 2 and 3, respectively), in accordance with the Safety Evaluation (Reference 3) to request an alternative from t he Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. In Enclosures 1 and 2 SCE identified 2022 for Units 2 and 3 as the year in which future inspection of the Examination Categories B-A and B-D RPV welds will be performed. The intent of this relief request (ISI-3-34) is to allow deferra l of the subject exami nations to the same time (2022) as the Examination Categor ies B-A and B-D RPV welds described in Enclosures 1 and 2.

During the ten-year ISI of the RPV shell, lower head, and nozzle welds performed in 2002 for Unit 2 and 2003 for Unit 3, SONGS also performed visual examinations of the

RPV interior attachments and the core s upport structure. Since the core support structure (called a core barrel on Combusti on Engineering designed plants) requires removal to facilitate exami nation of the RPV shell, lower head, and nozzle welds, the visual examinations of ASME Exami nation Categories B-N-2 and B-N-3 have historically been performed during the same outage at the end of the ISI interval.

Performing all core barrel removed related examinations during the same refueling outage will result in significant savings in dose and outage duration since the same equipment and personnel used for visual and vo lumetric examinatio n of the RPV shell welds and nozzle welds from the RPV interi or can be used to implement the required RVI examinations. Additionally, removing the RPV internals only once to accommodate all the examinations discussed in this relief request would result in significant savings in radiation exposure.

Proposed Alternative and Basis for Use The current SONGS Unit 2 and Unit 3 third 10-year ISI interval began on August 18, 2003 and is scheduled to end on August 17, 2013. An extension of one year is allowed in ASME, Sect ion XI, IWA-2430(d)(1).

SONGS proposes to perform the subject examinations during the f ourth ten-year ISI interval for both Units 2 and 3. The subj ect examinations woul d need to be performed during the 1st quarter 2012 refueling outage for Unit 2 and the 4th quarter 2012 outage for Unit 3, pending approval of this relief request. The proposed alternative inspection would enable the subject examinations to be performed during refueling outages in 2022 with the risk-informed extens ion of the reactor vessel ISI. In accordance with 10 CFR 50.55a(a)(3)(ii), this interval extension is requested on the basis that performing the examination of the RPV interior attachments and core support structure separate in time from the RPV shell, head, and nozzle welds would result in hardship or unusual difficulty without a compensating increase in quality or safety.

San Onofre Units 2 and 3 10 CFR 50.55a Request ISI-3-34 In Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship without Compensating Increase in the Level of Quality and Safety Page 3 of 4 The full scope examination required by ASME Examination Categories B-N-2 and B-N-3 requires the removal of all the fuel and the core barrel from the RPV. An unnecessary risk is created by removal of the core barrel to perform a visual examination without a compensating increase in quality or safety. Further, the radiation exposure to estab lish the conditions for and pe rform the ASME Examination Categories B-N-2 and B-N-3 exam inations would essentially double if the subject examinations were performed separate in ti me from the RPV shell, lower head, and nozzle weld examinations.

The visual examinations of the RPV interior attachments and the core support structure have been performed twice per unit at SONGS.

During the second ten-year ISI interval visual inspection on Unit 3, three indications were noted. These indications were evaluated as non-relev ant conditions and acceptable for continued operation. There were no other relevant indications noted during the examinations. The examinations were last performed during the 2002 refueling outage for Unit 2 and the 2003 outage for Unit 3 with acceptable result

s. Additionally, review of industry surveys indicate that these examinations have been performed many times by the industry without any significant findings relevant to the SONGS r eactor vessel design.

As stated in Reference 2, "-it must be recognized that all reactor coolant pressure boundary failures occurring to date have been ident ified as a result of leakage, and were discovered by visual examination.

The proposed RV ISI interval extension does not alter the visual examination interval. The reactor vessel would undergo, as a minimum, the Section XI Examination Ca tegory B-P pressure tests and visual examinations conducted at the end of each re fueling before plant start-up, as well as leak tests with visual examinations t hat precede each start-up following maintenance or repair activities."

The minimum visual examinations discussed in Reference 2 are not the subject examinations (i.e., B-N-2 and B-N-3) of this relief request. During the 2012 refueling outages for Units 2 and 3, SONGS will be performing the ASME Examination Category B-N-1 visual examination. This examination wil l include the space that is made accessible for exami nation by the removal of components during normal refueling outages. This exami nation is required onc e each period and will provide reasonable assurance of structural integrity. As discussed further in Reference 2, defenses against human errors are preserved with the increase in inspection interval. Specifically, the increase in the inspection interval reduces the frequency for which the reactor vessel lowe r internals need to be removed thereby reducing the possibility for human error and damage to the core.

Therefore, in accordance wit h 10 CFR 50.55a(a)(3)(ii), this interval change from 10 to 20 years for the subject examinations is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

San Onofre Units 2 and 3 10 CFR 50.55a Request ISI-3-34 In Accordance with 10 CFR 50.55a(a)(3)(ii) Hardship without Compensating Increase in the Level of Quality and Safety Page 4 of 4 Duration of Proposed Alternative This proposed alternative is applicable to the third and fourth t en-year ISI for the Examination Categories B-N-2 and B-N-3, Item Numbers B13.50, B13.60, and B13.70 visual examinations.

Precedents:

Letter from M. D. Flaherty , (Calvert Cliffs Nuclear Power Plant) to Document Control Desk (NRC), dated February 18, 2009 (ADAMS Accession No.

ML090540062);

Subject:

"Unit 1; Docket No. 50-317, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)

Letter from N. L. Salgado (NRC) to J. A. Spina (Calvert Cliffs Nuclear Power Plant, dated November 9, 2009 (ADAMS Acce ssion No. ML093030052);

Subject:

"Relief Requests Inservice Inspection (ISI)-022 and ISI-023 to Extend Reactor Vessel and

Reactor Vessel Internal Weld Examinations - Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC Nos. ME0668 and ME0669)

References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition with the 1996 Addenda 2. WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval, June 2008 3. Final Safety Evaluation For Pre ssurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revi sion 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval" (TAC No.

MC9768), Dated May 8, 2008