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| | issue date = 05/11/2010 | | | issue date = 05/11/2010 |
| | title = Relief Request 13R-06 for Detailed Visual Examination During Appendix J Pnuematic Leakage Testing | | | title = Relief Request 13R-06 for Detailed Visual Examination During Appendix J Pnuematic Leakage Testing |
| | author name = Campbell S J | | | author name = Campbell S |
| | author affiliation = NRC/NRR/DORL/LPLIII-2 | | | author affiliation = NRC/NRR/DORL/LPLIII-2 |
| | addressee name = Pardee C G | | | addressee name = Pardee C |
| | addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear | | | addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear |
| | docket = 05000456, 05000457 | | | docket = 05000456, 05000457 |
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| ==Dear Mr. Pardee:== | | ==Dear Mr. Pardee:== |
| By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML09251 0267), Exelon Generation Company, LLC (the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 10-year lSI interval for Braidwood is currently scheduled to end on July 28,2018, for Unit 1 and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code, Section XI, requirements in Subarticle 5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed the licensee's submittal and has determined that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, and that the alternative proposed in RR 13R-06 will provide reasonable assurance of structural integrity and leak tightness of the affected components. | | By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267), Exelon Generation Company, LLC (the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 10-year lSI interval for Braidwood is currently scheduled to end on July 28,2018, for Unit 1 and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code, Section XI, requirements in Subarticle 5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed the licensee's submittal and has determined that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, and that the alternative proposed in RR 13R-06 will provide reasonable assurance of structural integrity and leak tightness of the affected components. |
| Therefore, the alternative proposed in RR 13-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), for the third 1 O-year lSI interval for Braidwood, Units 1 and 2. All other ASME Code, Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector. | | Therefore, the alternative proposed in RR 13-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), for the third 1 O-year lSI interval for Braidwood, Units 1 and 2. All other ASME Code, Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector. |
| The NRC staffs safety evaluation is enclosed. | | The NRC staffs safety evaluation is enclosed. |
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| ==1.0 INTRODUCTION== | | ==1.0 INTRODUCTION== |
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| By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML09251 0267, Reference 1), Exelon Generation Company, LLC (EGC, the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 1 O-year lSI interval for Braidwood is currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code, Section XI, requirements in Subarticle IWE-5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed and evaluated the information provided by the licensee in its submittal. | | By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267, Reference 1), Exelon Generation Company, LLC (EGC, the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 1 O-year lSI interval for Braidwood is currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code, Section XI, requirements in Subarticle IWE-5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed and evaluated the information provided by the licensee in its submittal. |
| The results of the NRC staff's review are presented in the remainder of this safety evaluation. | | The results of the NRC staff's review are presented in the remainder of this safety evaluation. |
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| Based on the information provided in Reference 1 related to RR 13R-06, and on the NRC staff's evaluation discussed above, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. All other ASME Code, Section XI, requirements for which relief was not specifically requested and approved in the subject RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. | | Based on the information provided in Reference 1 related to RR 13R-06, and on the NRC staff's evaluation discussed above, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. All other ASME Code, Section XI, requirements for which relief was not specifically requested and approved in the subject RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. |
| -REFERENCES Letter No. RS-09-118 dated September 4, 2009, from Patrick R. Simpson, EGC, to USNRC, "Submittal of Relief Requests Associated with the Third Inservice Inspection Interval" (ADAMS Accession No. ML09251 0267). Letter dated September 22, 2008, from L. James (NRC/NRR) to R. L. Anderson (Duane Arnold Energy Center), "Duane Arnold Energy Center Safety Evaluation for Relief Requests MC-R001 and MC-P001 for Second Containment Inservice Inspection Interval" (ADAMS Accession No. ML082460235). | | -REFERENCES Letter No. RS-09-118 dated September 4, 2009, from Patrick R. Simpson, EGC, to USNRC, "Submittal of Relief Requests Associated with the Third Inservice Inspection Interval" (ADAMS Accession No. ML092510267). Letter dated September 22, 2008, from L. James (NRC/NRR) to R. L. Anderson (Duane Arnold Energy Center), "Duane Arnold Energy Center Safety Evaluation for Relief Requests MC-R001 and MC-P001 for Second Containment Inservice Inspection Interval" (ADAMS Accession No. ML082460235). |
| Principal Dan Hoang, NRR William Jessup, NRR May 11, 2010 C. Pardee Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this action. Docket No. STN Safety cc w/encl: Distribution via DISTRIBUTION: | | Principal Dan Hoang, NRR William Jessup, NRR May 11, 2010 C. Pardee Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this action. Docket No. STN Safety cc w/encl: Distribution via DISTRIBUTION: |
| PUBLIC RidsNrrPMBraidwood Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource Sincerely, IRA! Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPL3-2 RlF RidsNrrDorlLpl3-2 Resource RidsNrrLATHarris Resource DHoang, NRR RidsNrrDeMceb Resource WJessup. NRR RidsOgcRp Resource RidsNrrDorlDpr Resource "'SE memo date NRR-028 OFFICE -2/PM LPL3-2/LA DE/MCEB/BC'" LPL3-21BC NAME HLindsay MDavid THarris MKhanna SCampbell 1/10 DATE 5/4/10 5/5/10 5/3/10 1/6/10 OFFICIAL RECORD COpy}} | | PUBLIC RidsNrrPMBraidwood Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource Sincerely, IRA! Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPL3-2 RlF RidsNrrDorlLpl3-2 Resource RidsNrrLATHarris Resource DHoang, NRR RidsNrrDeMceb Resource WJessup. NRR RidsOgcRp Resource RidsNrrDorlDpr Resource "'SE memo date NRR-028 OFFICE -2/PM LPL3-2/LA DE/MCEB/BC'" LPL3-21BC NAME HLindsay MDavid THarris MKhanna SCampbell 1/10 DATE 5/4/10 5/5/10 5/3/10 1/6/10 OFFICIAL RECORD COpy}} |
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Category:Code Relief or Alternative
MONTHYEARML22307A2462022-11-10010 November 2022 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (EPIDs L-2021-LLR-0035 and L-2021-LLR-0036) RS-22-110, Supplemental Information - Proposed Alternatives Related to the Steam Generators and Request.2022-09-20020 September 2022 Supplemental Information - Proposed Alternatives Related to the Steam Generators and Request. RS-22-036, Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2022-03-10010 March 2022 Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds ML21280A0782021-10-27027 October 2021 Proposed Alternative to Use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-893 (Epids L-2020-LLR-0147 and L-2020-LLR-0148) ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21134A0062021-08-0303 August 2021 Proposed Alternatives to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (Epids L-2020-LLR-0099 and L-2020-LLR-0100) RS-21-008, Request for Alternative: One-Time Deferral of Follow-Up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2)2021-01-25025 January 2021 Request for Alternative: One-Time Deferral of Follow-Up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 ML20188A2642020-07-0606 July 2020 Clinton Power Station, R.E. Ginna Station, Limerick Station, Nine Mile Point Station & Peach Bottom Station - Proposed Alternative to Utilize Code Case OMN-26 - Response to Request for Additional Information ML20113F0412020-04-30030 April 2020 Proposed Alternative to the Submittal Schedule for Certain Reports (COVID-19) ML20099D9552020-04-17017 April 2020 Request to Use Provisions in the 2013 Edition of the ASME Boiler and Pressure Vessel Code for Performing Non-Destructive Examinations RS-20-020, Relief Request for Alternative Frequency to Supplemental Indication Requirements of 10 CFR 50.55a(b)(3)(xi)2020-02-28028 February 2020 Relief Request for Alternative Frequency to Supplemental Indication Requirements of 10 CFR 50.55a(b)(3)(xi) ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19155A0602019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19141A0202019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19136A3862019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML19161A2572019-06-0404 June 2019 BWR Fleet Msv/Srv - Testing Frequency Relief Request NRC Pre-Application Meeting June 4, 2019 JAFP-19-0023, Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds2019-02-15015 February 2019 Relief Request Associated with the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography for Ferritic and Austenitic Piping Welds RS-19-015, Relief Request 14-RS-1 to Implement Code Case OMN-132019-01-31031 January 2019 Relief Request 14-RS-1 to Implement Code Case OMN-13 ML18347B4192019-01-17017 January 2019 Relief Request I4R-03, Relief from ASME Requirements for the Fourth 10-Year Inservice Inspection Interval Related to Degraded Canopy Seal Welds Associated with Control Rod Drive Mechanism ML18318A3342019-01-17017 January 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML18331A0372019-01-17017 January 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code ML18305A3602018-12-0606 December 2018 Relief from the Requirements of the American Society of Mechanical Engineers Code RS-18-125, Proposed Alternative Requirements for the Repair and Examination of Reactor Pressure Vessel Head Penetration Nozzles for the Fourth Lnservice Inspection Interval in Accordance with 10 CFR 50.55a(z)(1)2018-10-11011 October 2018 Proposed Alternative Requirements for the Repair and Examination of Reactor Pressure Vessel Head Penetration Nozzles for the Fourth Lnservice Inspection Interval in Accordance with 10 CFR 50.55a(z)(1) RS-18-123, Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2)2018-09-24024 September 2018 Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) JAFP-18-0052, Proposed Alternative to Utilize Code Case N-8792018-05-30030 May 2018 Proposed Alternative to Utilize Code Case N-879 JAFP-18-0053, Proposed Alternative to Utilize Code Cases N-878 and N-8802018-05-30030 May 2018 Proposed Alternative to Utilize Code Cases N-878 and N-880 RS-17-168, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1)2017-12-20020 December 2017 Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1) RS-17-168, Braidwood Station, Unit 2, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1)2017-12-20020 December 2017 Braidwood Station, Unit 2, Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(1) ML17249A2982017-11-13013 November 2017 Relief from the Requirements of the ASME Code (CAC No. MF9597; EPID L-2017-LLR-0021) ML17170A0132017-06-26026 June 2017 Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548) ML17095A2682017-03-31031 March 2017 Submittal of Request for Relief for Extension of Examination Interval for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces: Attachment 1, Relief Request ML17054C2552017-03-15015 March 2017 Request for Relief from Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) ML16230A2372016-09-0606 September 2016 Fleet Request for Proposed Alternative to Use ASME Code Case N-513-4 (CAC Nos. MF7301-MF7322) ML16162A2112016-06-29029 June 2016 Request for Use of Alternative ML16109A3372016-04-27027 April 2016 Relief from the Requirements of the ASME Code (CAC Nos. MF6715, MF6716, MF6717, and MF6718) ML14303A5062014-12-10010 December 2014 Relief from the Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval ML13016A5152013-01-30030 January 2013 Relief from the Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection ML12121A6372012-05-10010 May 2012 Request to Use Code Case N-789 ML12108A1232012-04-19019 April 2012 Safety Evaluation in Support of the Third 10-Year Inservice Inspection Interval Request for Relief 13R-08 (Tac Nos. ME6024 and ME6025) ML1113306532011-06-0606 June 2011 Unacceptable with Opportunity to Supplement Alt. to ASME Code Requirements for Repair of Reactor Vessel Head Penetrations (TACs ME6071, ME6072, ME6073, and ME6074) ML1105909212011-03-0303 March 2011 Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds ML1011905272010-05-11011 May 2010 Relief Request 13R-06 for Detailed Visual Examination During Appendix J Pnuematic Leakage Testing ML1012301792010-05-10010 May 2010 Relief Request I3R-03 for Examination of Structural Weld Overlays ML1008504952010-03-26026 March 2010 Application Accepted - Braidwood & Byron Relief Request Re. ASME Code Case N-729-1 (TACs ME3510 - ME3513) RS-10-046, Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds2010-03-12012 March 2010 Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds ML1004806992010-03-0808 March 2010 Relief Request 12R-50 for Second 10-Year Inservice Inspection Interval ML0930702712009-11-0505 November 2009 Risk-Informed Relief Request 13R-01 for Certain Pressure Retaining Piping Welds ML0803701812008-04-14014 April 2008 Relief Request RV-1 for the Third Interval Inservice Testing Program 2022-09-20
[Table view] Category:Letter
MONTHYEARIR 05000456/20230042024-02-0202 February 2024 Integrated Inspection Report 05000456/2023004 and 05000457/2023004 ML24025C7242024-01-29029 January 2024 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000456/2024002; 05000457/2024002 IR 05000457/20230112024-01-25025 January 2024 2B Auxiliary Feedwater Pump Diesel Fuel Oil Dilution Report 05000457/2023011 and Preliminary Greater than Green Finding and Apparent Violation ML24018A0362024-01-17017 January 2024 Paragon Energy Solutions, Defect with Detroit Diesel/Mtu Fuel Injectors P/N R5229660 Cat Id 0001390618 RS-24-004, Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators2024-01-11011 January 2024 Proposed Alternative to the Distribution Requirements of ASME Code Table IWC-2411-1 for the Steam Generators ML23348A2162023-12-15015 December 2023 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0030 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000456/20200232023-12-15015 December 2023 Baseline Security Inspection Document; 05000456/2023/402; 05000457/2023/402 ML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23339A0452023-12-0505 December 2023 Request for Information for an NRC Post-Approval Site Inspection for License Renewal Inspection Report 05000546/2024010 ML23313A1552023-12-0101 December 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23331A8922023-11-22022 November 2023 Supplement - Braidwood Security Rule Exemption Request ISFSI Docket No. Reference 05000457/LER-2023-001, Submittal of LER 2023-001-00 for Braidwood Station, Unit 2, Train B Auxiliary Feedwater Pump Was Inoperable Due to Degraded Oil in the Crank Case2023-11-17017 November 2023 Submittal of LER 2023-001-00 for Braidwood Station, Unit 2, Train B Auxiliary Feedwater Pump Was Inoperable Due to Degraded Oil in the Crank Case ML23321A0442023-11-17017 November 2023 Notification of Deviation from Electric Power Research Institute (EPRI) Topical Report MRP-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline RS-23-118, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums RS-23-114, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2023-11-0101 November 2023 Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds IR 05000456/20234012023-10-18018 October 2023 Security Baseline Inspection Report 05000456/2023401 and 05000457/2023401 IR 05000456/20230102023-10-18018 October 2023 Functional Engineering Inspection Commercial Grade Dedication Report 05000456/2023010 and 05000457/2023010 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans RS-23-108, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles2023-10-11011 October 2023 Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles RS-23-105, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2023-10-10010 October 2023 Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections RS-23-093, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.2023-09-29029 September 2023 License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel. ML23226A0062023-09-19019 September 2023 Review of License Renewal Commitment Number 10 Submittal ML23180A1692023-09-11011 September 2023 Calvert Cliff Units 1 & 2, and R.E. Ginna Plant - Withdrawal of Proposed Alternatives to American Society of Mechanical Engineers (ASME) Requirements (Epids L-2022-LRR-0074, 0076, 0079, 0091, 0092, 0093 and 0094) IR 05000456/20230052023-08-30030 August 2023 Updated Inspection Plan for Braidwood Station Report 05000456/2023005 and 05000457/2023005 ML23234A2462023-08-25025 August 2023 Confirmation of Initial License Examination IR 05000456/20230022023-08-0303 August 2023 Integrated Inspection Report 05000456/2023002 and 05000457/2023002 ML23188A1292023-07-26026 July 2023 Issuance of Amendment Nos. 233 and 233 Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1 ML23087A0762023-07-13013 July 2023 Issuance of Amendment Nos. 232 and 232 Revision of Technical Specifications for the Ultimate Heat Sink ML23191A8442023-07-10010 July 2023 05000456; 05000457 Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III RS-23-083, Withdrawal - Proposed Alternatives Related to the Steam Generators2023-06-27027 June 2023 Withdrawal - Proposed Alternatives Related to the Steam Generators RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23110A1152023-06-12012 June 2023 Environmental Assessment and Finding of No Significant Impact Related to a Requested Increase in Ultimate Heat Sink Temperature (EPID L-2023-LLA-0042) (Letter) RS-23-074, Supplement to Application for License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2023-06-0909 June 2023 Supplement to Application for License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-23-075, Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process2023-06-0707 June 2023 Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process RS-23-050, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube.2023-05-22022 May 2023 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube. ML23138A1342023-05-18018 May 2023 Information Meeting with a Question and Answer Session to Discuss NRC 2022 End-Of-Cycle Plant Performance Assessment of Braidwood Station and Byron Station ML23132A0472023-05-12012 May 2023 Submittal of 2022 Annual Radiological Environmental Operating Report ML23130A0072023-05-10010 May 2023 Submittal of Core Operating Limits Report Cycle 24, Rev. 16 IR 05000456/20230012023-05-0808 May 2023 Integrated Inspection Report 05000456/2023001 and 05000457/2023001 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML23118A0202023-04-28028 April 2023 Submittal of 2022 Annual Radioactive Effluent Release Report ML23110A3202023-04-21021 April 2023 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection RS-23-056, Response to Request for Additional Information to Braidwood Station, Unit 1, and Byron Station, Unit 1, for Steam Generator License Renewal Response to Commitment 102023-04-20020 April 2023 Response to Request for Additional Information to Braidwood Station, Unit 1, and Byron Station, Unit 1, for Steam Generator License Renewal Response to Commitment 10 RS-23-055, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2023-04-10010 April 2023 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML23095A1292023-04-0505 April 2023 Steam Generator Tube Inspection Report for Refueling Outage 23 ML23094A1352023-04-0404 April 2023 Request for Information for Nrc Commercial Grade Dedication Inspection Inspection Report 05000456/2023010 05000457/2023010 RS-23-052, License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink2023-03-24024 March 2023 License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, Ultimate Heat Sink RS-23-049, Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-23023 March 2023 Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations 2024-02-02
[Table view] Category:Safety Evaluation
MONTHYEARML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23241A9092023-09-19019 September 2023 Enclosure 2 - Non-Proprietary - Review of License Renewal Commitment Number 10 Safety Evaluation ML23188A1292023-07-26026 July 2023 Issuance of Amendment Nos. 233 and 233 Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1 ML23087A0762023-07-13013 July 2023 Issuance of Amendment Nos. 232 and 232 Revision of Technical Specifications for the Ultimate Heat Sink ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML22364A0242023-03-0101 March 2023 R. E. Ginna Nuclear Power Plant Issuance of Amendments Nos. 231, 231, 232, 232, and 154 Regarding Adoption of TSTF-246 ML22293B8052022-11-30030 November 2022 Constellation Energy Generation, Llc_Fleet - Request to Authorize Use Honeywell Mururoa V4F1 R Supplied Air Suits ML22307A2462022-11-10010 November 2022 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (EPIDs L-2021-LLR-0035 and L-2021-LLR-0036) ML22210A0312022-08-30030 August 2022 Issuance of Amendments Nos. 230, 230, 230, and 230, Respectively, Regarding Adoption of Technical Specifications Task Force Traveler (TSTF) 501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control ML22173A1812022-08-11011 August 2022 Issuance of Amendment No. 229 to Remove License Condition ML22173A2142022-08-10010 August 2022 Issuance of Amendments Nos. 228 and 228 Revision of Technical Specifications for the Ultimate Heat Sink ML22095A2702022-05-12012 May 2022 Issuance of Amendment Nos. 227, 227, 229, 229, and 245, Respectively, Regarding Adoption of TSTF 273, Safety Function Determination Program Clarifications ML22090A0862022-04-29029 April 2022 Amendments to Adopt TSTF-541,Rev.2,Add Exceptions to Surveillance Requirements for Valves,Dampers Locked in Actuated Position ML22026A4892022-03-22022 March 2022 Issuance of Amendment Nos. 225, 225, 227, 227, and 148, Respectively, Regarding Issues Identified in Westinghouse Documents (EPID L-2021-LLA-0066) Nonproprietary ML21347A0382022-01-13013 January 2022 Issuance of Amendments to Revise Reactor Coolant Leakage Requirements ML21277A2482021-11-16016 November 2021 Letter with Enclosure 4, Safety Evaluation for Transfer of Licenses and Draft Conforming License Amendments (Public Version) ML21280A0782021-10-27027 October 2021 Proposed Alternative to Use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-893 (Epids L-2020-LLR-0147 and L-2020-LLR-0148) ML21230A2062021-09-0303 September 2021 Proposed Alternative to Use ASME OM Code Case OMN-28 ML21216A2202021-08-0505 August 2021 Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting ML21134A0062021-08-0303 August 2021 Proposed Alternatives to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (Epids L-2020-LLR-0099 and L-2020-LLR-0100) ML21154A0462021-07-13013 July 2021 Issuance of Amendments Nos. 222 and 222 Revision of Technical Specifications for the Ultimate Heat Sink ML21166A1682021-06-25025 June 2021 ML21060B2812021-04-0202 April 2021 Issuance of Amendments Nos. 221, 221, 224, and 224 Regarding Technical Specifications 3.8.1, AC Sources-Operating ML21054A0082021-03-10010 March 2021 Issuance of Amendment Nos. 220 and 220 One-Time Deferral of Steam Generator Tube Inspections ML21063A0162021-03-0808 March 2021 Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Alternative to 10 CFR 50.55a(z)(2) ML21039A6362021-02-17017 February 2021 R. E. Ginna - Proposed Alternative to Use the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-885 ML21028A6732021-02-0303 February 2021 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L-2020-LLR-011 ML20317A0012020-12-28028 December 2020 Non-Proprietary, Issuance of Amendment Nos. 219, 219, 223, and 223, Revise Loss-of-Coolant Accident Methodology in TS 5.6.5, Core Operating Limits Report (COLR) ML20287A1302020-11-0505 November 2020 Review of Quality Assurance Program Changes ML20269A2002020-09-30030 September 2020 Request to Use a Provision of Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI (EPID L2020-LLR-0117 ML20245E4192020-09-24024 September 2020 Issuance of Amendments Nos. 218 and 218 Revision of Technical Specifications for the Ultimate Heat Sink ML20163A0462020-09-18018 September 2020 Issuance of Amendments Nos. 217, 217, 221, and 221, Revise Technical Specification 5.6.6 to Allow Use of Areva Np Topical Report BAW-2308 ML20167A0072020-09-11011 September 2020 R. E. Ginna - Issuance of Amendment Nos. 216, 216, 220, 220, and 143 - Adoption of TSTF-567, Rev. 1, Add Containment Sump Technical Specifications to Address GSI-191 Issues ML20149K6982020-09-10010 September 2020 Issuance of Amendment Nos. 215, 215, 219, and 219 Permanent Extension of Type a and Type C Leak Rate Test Frequencies ML20232A1712020-09-0101 September 2020 Request to Use Alternative Code Case OMN-26 ML20153A8042020-07-31031 July 2020 Co. - Issuance of Amendments Revising Emergency Action RA3 ML20141L6362020-07-10010 July 2020 Issuance of Amendments Based on TSTF-427,Allowance for Nontechnical Specification Barrier Degradation on Supported System Operability,Rev 2 ML20134H9402020-07-0808 July 2020 Issuance of Amendments Revising the High Radiation Area Administrative Controls ML20118C4292020-06-0909 June 2020 Issuance of Amendments Revision of Technical Specifications for the Ultimate Heat Sink ML20133K0932020-05-14014 May 2020 Relief from the Requirements of the ASME Code ML20111A0002020-05-0101 May 2020 Issuance of Amendment No. 209, Revision Technical Specification 5.5.9, Steam Generator (SG) Program, for One-Time Revision to Frequency of SG Tube Inspections (Exigent Circumstances) ML20113F0412020-04-30030 April 2020 Proposed Alternative to the Submittal Schedule for Certain Reports (COVID-19) ML20021A0702020-04-0606 April 2020 Issuance of Amendments to Delete License Conditions for Decommissioning Trusts ML19331A7252020-02-14014 February 2020 Issuance of Amendments Revising Emergency Action Levels ML20028E3992020-02-0404 February 2020 Proposed Alternative to Use ASME Code Case N-879 ML19269C5342019-09-27027 September 2019 Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques ML19240B1122019-09-0909 September 2019 Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Code ML19190A0812019-08-28028 August 2019 Issuance of Amendments Regarding Limiting Condition of Operation for Inoperability of Snubbers ML19192A2442019-07-18018 July 2019 Proposed Alternative to Use ASME Code Cases N-878 and N-880 ML19136A3862019-06-0505 June 2019 Relief from the Requirements of the American Society of Mechanical Engineers Code 2023-09-19
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 11,2010 Mr. Charles Pardee President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 BRAIDWOOD STATION, UNITS 1 AND 2 -RELIEF REQUEST 13R-06 FOR DETAILED VISUAL EXAMINATION DURING APPENDIX J PNUEMATIC LEAKAGE TESTING (TAC NOS. ME2194 AND ME2195)
Dear Mr. Pardee:
By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267), Exelon Generation Company, LLC (the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 10-year lSI interval for Braidwood is currently scheduled to end on July 28,2018, for Unit 1 and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code,Section XI, requirements in Subarticle 5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed the licensee's submittal and has determined that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, and that the alternative proposed in RR 13R-06 will provide reasonable assurance of structural integrity and leak tightness of the affected components.
Therefore, the alternative proposed in RR 13-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), for the third 1 O-year lSI interval for Braidwood, Units 1 and 2. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
The NRC staffs safety evaluation is enclosed.
C. Pardee -2 Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this Sincerely, Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. STN 50-456 and 50-457 Safety cc w/encl: Distribution via UNITED NUCLEAR REGULATORY WASHINGTON,1l.C.-20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR RELIEF REQUEST NO. EXELON GENERATION COMPANY, BRAIDWOOD STATION, UNITS 1 AND DOCKET NOS. STN 50-456 AND
1.0 INTRODUCTION
By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267, Reference 1), Exelon Generation Company, LLC (EGC, the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 1 O-year lSI interval for Braidwood is currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code,Section XI, requirements in Subarticle IWE-5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Cooled Power Reactors," pneumatic leakage testing. The NRC staff has reviewed and evaluated the information provided by the licensee in its submittal.
The results of the NRC staff's review are presented in the remainder of this safety evaluation.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.55a, "Code and standards," address the use of codes and standards as they relate to structures, systems, and components, which must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.
Specifically, 10 CFR 50.55a(b)(2)(ix) identifies the regulatory conditions that apply to the use of Subsection IWE of the ASME Code for the examination of metal containments and the liners of concrete containments, while 10 CFR 50.55a(g) delineates the requirements for lSI of those components (including supports), which are classified as ASME Code Class 1, Class 2, or Class 3. According to 10 CFR 50.55a(a)(3), proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g) and (h) of 10 CFR 50.55a, or portions thereof, may be used when authorized by NRC's Director of the Office of Nuclear Reactor Regulation provided the applicant demonstrates that: (i) the proposed
-2 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 RR 13R-06, the licensee requested relief from certain ASME Code,Section XI, Subsection IWE requirements pursuant to 10 CFR 50.55a(a)(3)(ii).
On September 22, 2008, the NRC staff authorized a similar relief request relating to the detailed visual examination requirements of Subsection IWE of the ASME Code for the second containment lSI Interval for the Duane Arnold Energy Center (ADAMS Accession No. I\IIL082460235, Reference 2). 3.0 TECHNICAL EVALUATION 3.1 Description of RR 13R-06 In Section 4.0 of RR 13R-06 in Reference 1, the licensee requested relief, pursuant to 10 CFR 50.55a(a)(3)(ii), from the requirements of Subarticle IWE-5240 to perform a detailed visual examination during the Appendix J local leak rate test for replacements (installed by mechanical connection) or minor repair activities. This request was submitted 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. In the RR, the licensee stated that, per IWE-2310(d), detailed visual examinations must be performed in accordance with Subarticle IWE-5240 to assess the structural condition of areas affected by repair/replacement activities.
The licensee also stated that the majority of the repair/replacement activities for Category E-A components at Braidwood have been associated with the replacement of bolted components (e.g., containment personnel
[air lock] components, containment penetration blind flange bolting, etc.). The licensee further stated that the existing components have been replaced either to improve reliability, because existing components were at the end of service life, or because existing material was lost or damaged during disassembly.
Class MC replacements or repairs at Braidwood have not been required due to degradation associated with the conditions described in Subarticie IWE-1241 , which would categorize components as augmented areas. The licensee also stated that, in accordance with site repair/replacement plans, when a Class MC Section XI repair or replacement is to be performed, construction code non-destructive examinations and pre-service examinations of replacement items are performed following installation of minor repairs or prior reassembly of the component, when access to surfaces is not limited, permitting the examiner full access to the existing and replacement materials requiring the pre-service inspection.
The licensee stated that the appropriate time to assess the structural condition of these locations is during component disassembly.
Components and connections are also inspected by mechanics as standard practice, whether or not the component is scheduled for an ASME Section XI periodic inspection.
3.1.1 Component Identification for this RR Code Class: MC Code
References:
IWE-2000 Examination and Inspection IWE-5000 System Pressure Tests Examination Categories:
E-Aand E-C Item Numbers: E1.10 and E4.10
Description:
IWE Components Subject to Repair/Replacement Pressure Testing Applicable Code Edition and Addenda The applicable Code edition and addenda relating to this RR are the ASME Code,Section XI, 2001 Edition through the 2003 Addenda. Applicable Code Requirement from which Relief is Requested ASME Code,Section XI (2001 Edition, 2003 Addenda), Subarticle IWE-S240, requires a detailed visual examination (per Subarticle IWE-2310) to be performed on areas affected by repairlreplacement activities during the post-repair/replacement pressure test required by Subarticle IWE-S220. Subarticle IWE-2310 states that detailed visual examinations shall be performed in accordance with IWE-2S00 and Table IWE-2S00-1, Examination Category E-A and E-C. The regulations at 10 CFR SO.SSa(b)(2)(ix)(G) require this detailed visual examination to be conducted using the VT-1 method. Subarticle IWE-S221 requires all repairlreplacement activities (except those noted in Subarticle IWE-S222) performed on the pressure retaining boundary of Class MC or Class CC components to be subjected to pneumatic leakage testing in accordance with 10 CFR Part SO, Appendix J, Paragraph IV.A. Subarticle IWE-S222 states that leakage tests for the following minor repair/replacement activities performed on the pressure retaining boundary may be deferred until the next scheduled leakage test, provided that nondestructive examination is performed in accordance with the station Repair/Replacement Program and Plan. Minor repairs include the following activities: Welds of attachments to the surface of the pressure retaining boundary Weld cavities, the depth of which does not penetrate the required design wall thickness by more than 10 percent Welds attaching penetrations that are NPS 1 [nominal pipe size of 1 inch] or smaller Subarticle IWE-S240 requires a detailed visual examination (per Subarticle IWE-231 0) on areas affected by repair/replacement activities that are performed during the Appendix J pneumatic leakage test. As noted above, 10 CFR SO.SSa{b)(2)(ix)(G) requires the VT-1 method to be used to conduct the examination in Item 4.11 (detailed visual) of Table IWE-2S00-1.
Additionally, 10 CFR 50.55a(b)(2)(ix)(H) requires the VT-3 method to be used to examine containment bolted connections and a subsequent VT-1 examination to be performed if any flaws or degradation are noted during the initial VT-3 examination.
3.1.4 Licensee's Basis for Request The licensee stated that performing a detailed visual examination (i.e., the VT-1 method per 10 CFR 50.55a(b)(2)(ix)(G>>
for replacements or minor repairs during the Appendix J pneumatic leakage test after IWE-2200 pre-service examinations are already performed, or after the components have been re-assembled, does not provide any additional assurance of safety. The licensee also stated that the conditions of interest for the detailed visual examinations described in Subarticle IWE-1241 would be readily apparent to the examiners and mechanics while the component was disassembled. The licensee further stated that depending on the unit's Technical Specification operating mode when the minor repair or replacement is performed, plant conditions could be such that the examiner performing the detailed visual examination during the Appendix J pneumatic leakage test would have to perform the examination in a neutron radiation field (e.g., personnel
[air lock] during unit operation), which would be contrary to the principles of ALARA [as low as is reasonably achievable]
for occupational radiation exposure.
The licensee stated that, depending on which Appendix J surveillance is required, access to the replaced component may not be possible.
For example, if an interlock barrel test is required, examiners cannot perform a meaningful examination because the area of interest cannot be entered, and access to the replaced component would be limited to looking through a sight glass (only present on the outer access door). The licensee also stated that at Braidwood, Appendix J pneumatic leakage testing is performed by operators using continuous use surveillances, along with calibrated equipment, and that performing a detailed visual examination in conjunction with the Appendix J surveillance does not provide any additional assurance of safety beyond the current Appendix J practices.
The licensee has, therefore, requested relief pursuant to 10 CFR 50.55a(a)(3)(ii), stating that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. 3.1.5 Proposed Alternative Examination and Duration In lieu of performing the detailed visual examination during the Appendix J pneumatic leakage tests after a replacement or minor repair at Braidwood, the licensee proposes to perform VT-1 examinations of Class MC or Class CC component repairs/replacements along with any other required IWE-2200 examinations.
The licensee stated that these examinations shall be performed in accordance with the Repair/Replacement Program prior to the conduct of the Appendix J pneumatic leakage test. The relief is requested for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2.
-5 3.2 NRC Staff Evaluation Per IWE-231 O(d), the purpose of the detailed visual examination (VT-1) required by Subarticle IWE-5240 is to assess the structural condition of areas affected by repair/replacement activities.
Subarticle IWE-5222 allows the leakage tests following the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c) to be deferred until the next scheduled leakage test, provided that the nondestructive examination is performed in accordance with the Repair/Replacement Program and Plan. With regards to the RR described herein, the Subarticle IWE-5240 requirements for the detailed visual examination (VT -1) during the pressure test will be completely met for the repairireplacement activities on the containment pressure boundary, except for the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c). In lieu of the Subarticle IWE-5240 detailed visual examination (VT-1) requirement for the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c), a non-destructive examination (such as RT, MT, etc.) of the affected area will be performed per the construction code or Repair/Replacement Program and Plan immediately following these minor repairireplacement activities for which relief is sought. Further, applicable leakage testing of the affected area will be performed in accordance with Appendix J. The NRC staff finds that the post-repair non-destructive examination of the affected area would satisfy the intent of IWE-231 O(d) for the detailed (VT -1) visual examination required by Subarticle IWE-5240 in assessing the structural condition of the area affected by minor repair/replacement activities.
Further, the applicable leakage test performed on the area affected by the minor repair/replacement, in accordance with the requirements and criteria in Appendix J would provide an acceptable measure of the structural integrity and leak tightness of the affected area. The NRC staff agrees that for minor repair/replacement activities, the VT-1 examination will not provide any additional assurance of safety beyond that provided by the nondestructive examination and leakage test. The NRC staff agrees that the presence of a VT-1 examiner in addition to the Appendix J test personnel would result in unnecessary additional dose exposure.
Based on the above, the NRC staff concludes that Braidwood, Units 1 and 2, complying with the specified requirement of Subarticle IWE-5240 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the areas affected by minor repair/replacement activities.
Therefore, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2.
4.0 CONCLUSION
Based on the information provided in Reference 1 related to RR 13R-06, and on the NRC staff's evaluation discussed above, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
-REFERENCES Letter No. RS-09-118 dated September 4, 2009, from Patrick R. Simpson, EGC, to USNRC, "Submittal of Relief Requests Associated with the Third Inservice Inspection Interval" (ADAMS Accession No. ML092510267). Letter dated September 22, 2008, from L. James (NRC/NRR) to R. L. Anderson (Duane Arnold Energy Center), "Duane Arnold Energy Center Safety Evaluation for Relief Requests MC-R001 and MC-P001 for Second Containment Inservice Inspection Interval" (ADAMS Accession No. ML082460235).
Principal Dan Hoang, NRR William Jessup, NRR May 11, 2010 C. Pardee Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this action. Docket No. STN Safety cc w/encl: Distribution via DISTRIBUTION:
PUBLIC RidsNrrPMBraidwood Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource Sincerely, IRA! Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation LPL3-2 RlF RidsNrrDorlLpl3-2 Resource RidsNrrLATHarris Resource DHoang, NRR RidsNrrDeMceb Resource WJessup. NRR RidsOgcRp Resource RidsNrrDorlDpr Resource "'SE memo date NRR-028 OFFICE -2/PM LPL3-2/LA DE/MCEB/BC'" LPL3-21BC NAME HLindsay MDavid THarris MKhanna SCampbell 1/10 DATE 5/4/10 5/5/10 5/3/10 1/6/10 OFFICIAL RECORD COpy