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{{#Wiki_filter:August 2, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Blvd. | {{#Wiki_filter:August 2, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Blvd. | ||
Glen Allen, VA 23060-6711 | Glen Allen, VA 23060-6711 | ||
==SUBJECT:== | ==SUBJECT:== | ||
MILLSTONE POWER STATION, UNIT NO. 2 - RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (EPID L-2020-LLR-0130) | MILLSTONE POWER STATION, UNIT NO. 2 - RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (EPID L-2020-LLR-0130) | ||
==Dear Mr. Stoddard:== | ==Dear Mr. Stoddard:== | ||
By {{letter dated|date=September 23, 2020|text=letter dated September 23, 2020}} (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20268A225), as supplemented by {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}} (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted relief requests (RRs) RR-04-31 and RR-04-32, which requested relief from the volumetric examination coverage requirements, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(5)(iii) on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry and materials of construction of the subject components for the Millstone Power Station, Unit No. 2 (Millstone 2). The relief is applicable to the third period of the fourth 10-year inservice inspection (ISI) interval for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | By {{letter dated|date=September 23, 2020|text=letter dated September 23, 2020}} (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20268A225), as supplemented by {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}} (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted relief requests (RRs) RR-04-31 and RR-04-32, which requested relief from the volumetric examination coverage requirements, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(5)(iii) on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry and materials of construction of the subject components for the Millstone Power Station, Unit No. 2 (Millstone 2). The relief is applicable to the third period of the fourth 10-year inservice inspection (ISI) interval for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | ||
In its {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}}, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage. | In its {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}}, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage. | ||
The U.S. Nuclear Regulatory Commission (NRC or Commission) staff has completed its review of the licensee's RR-04-32 request for Millstone 2. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff has determined that it is impractical for the licensee to comply with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, requirement; that the examinations performed to the extent practical provide reasonable assurance of structural integrity and leak tightness of the subject welds; and that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. | The U.S. Nuclear Regulatory Commission (NRC or Commission) staff has completed its review of the licensee's RR-04-32 request for Millstone 2. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff has determined that it is impractical for the licensee to comply with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, requirement; that the examinations performed to the extent practical provide reasonable assurance of structural integrity and leak tightness of the subject welds; and that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. | ||
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the staff grants relief for | Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the staff grants relief for | ||
D. Stoddard | D. Stoddard the subject examinations of the components contained in RR-04-32 for the fourth 10-year ISI interval at Millstone 2. | ||
All other ASME Code, Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector. | All other ASME Code, Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector. | ||
If you have any questions, please contact the Millstone project manager, Richard Guzman, at 301-415-1030 or by e-mail to Richard.Guzman@nrc.gov. | If you have any questions, please contact the Millstone project manager, Richard Guzman, at 301-415-1030 or by e-mail to Richard.Guzman@nrc.gov. | ||
Sincerely, | Sincerely, James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336 | ||
==Enclosure:== | ==Enclosure:== | ||
Safety Evaluation cc: Listserv James G. | |||
Danna Digitally signed by James G. Danna Date: 2021.08.02 11:19:18 -04'00' | |||
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATION PERFORMED IN THE FOURTH 10-YEAR INSPECTION INTERVAL DOMINION ENERGY NUCLEAR CONNECTICUT, INC. | |||
MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 | |||
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATION PERFORMED IN THE FOURTH 10-YEAR INSPECTION INTERVAL DOMINION ENERGY NUCLEAR CONNECTICUT, INC. | |||
MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 | |||
==1.0 INTRODUCTION== | |||
By {{letter dated|date=September 23, 2020|text=letter dated September 23, 2020}} (Agencywide Documents and Access Management System (ADAMS) Accession No. ML20268A225), as supplemented by {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}} (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc., (the licensee) submitted Relief Request (RR) RR-04-32 from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2004 Edition, under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) | By {{letter dated|date=September 23, 2020|text=letter dated September 23, 2020}} (Agencywide Documents and Access Management System (ADAMS) Accession No. ML20268A225), as supplemented by {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}} (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc., (the licensee) submitted Relief Request (RR) RR-04-32 from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2004 Edition, under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) | ||
Section 50.55a(g)(5)(iii), for limited coverage examinations performed in the third inspection period of the fourth 10-year inservice inspection (ISI) interval for Millstone Power Station, Unit No. 2 (Millstone 2). The fourth 10-year ISI interval began on April 1, 2010, and ended on March 31, 2020. | Section 50.55a(g)(5)(iii), for limited coverage examinations performed in the third inspection period of the fourth 10-year inservice inspection (ISI) interval for Millstone Power Station, Unit No. 2 (Millstone 2). The fourth 10-year ISI interval began on April 1, 2010, and ended on March 31, 2020. | ||
In its {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}}, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage. | In its {{letter dated|date=January 14, 2021|text=letter dated January 14, 2021}}, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage. | ||
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry, and materials of construction, of the subject components. | Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry, and materials of construction, of the subject components. | ||
==2.0 REGULATORY EVALUATION== | |||
Regulation 10 CFR 50.55a(g) requires that the ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with the ASME Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, and applicable editions and addenda. | Regulation 10 CFR 50.55a(g) requires that the ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with the ASME Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, and applicable editions and addenda. | ||
Regulation 10 CFR 50.55a(g)(4) mandates the use of the ASME Code, Section XI, which states, in part, that ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code, Section XI, Rules for Inservice Inspection of Nuclear | Regulation 10 CFR 50.55a(g)(4) mandates the use of the ASME Code, Section XI, which states, in part, that ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a) twelve months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b). | ||
Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a) twelve months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b). | |||
Regulation 10 CFR 50.55a(g)(5)(iii), requires that if a licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit information to support the determinations. | Regulation 10 CFR 50.55a(g)(5)(iii), requires that if a licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit information to support the determinations. | ||
Regulation 10 CFR 50.55a(g)(5)(iv), requires that where an examination requirement by the Code or Addenda is determined to be impractical by a licensee and is not included in the revised ISI program as permitted by paragraph 10 CFR 50.55a(g)(4), the basis for this determination must be demonstrated to the satisfaction of the Commission not later than 12 months after the expiration of the initial 120 month period of operation from the start of facility commercial operation and each subsequent 120 month period of operation during which the examination is determined to be impractical. | Regulation 10 CFR 50.55a(g)(5)(iv), requires that where an examination requirement by the Code or Addenda is determined to be impractical by a licensee and is not included in the revised ISI program as permitted by paragraph 10 CFR 50.55a(g)(4), the basis for this determination must be demonstrated to the satisfaction of the Commission not later than 12 months after the expiration of the initial 120 month period of operation from the start of facility commercial operation and each subsequent 120 month period of operation during which the examination is determined to be impractical. | ||
Regulation 10 CFR 50.55a(g)(6)(i) states in part that the Commission will evaluate determinations, under paragraph 10 CFR 50.55a(g)(5), that Code requirements are impractical. | Regulation 10 CFR 50.55a(g)(6)(i) states in part that the Commission will evaluate determinations, under paragraph 10 CFR 50.55a(g)(5), that Code requirements are impractical. | ||
The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law. | The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law. | ||
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the use of an alternative and the NRC to grant the relief requested by the licensee. | Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the use of an alternative and the NRC to grant the relief requested by the licensee. | ||
3.1 | ==3.0 TECHNICAL EVALUATION== | ||
3.1 Relief Request RR-04-32 ASME Code Components Affected The affected components are ASME Code Class 1 welds, BPV-C-5049B and BPV-C-5051B, as shown in the table below. In accordance with Table IWB-2500-1 of the ASME Code, Section XI, the subject welds are classified as Examination Category R-A, Risk-Informed Piping Examinations, Item No. R1.20, Elements not Subject to a Degradation Mechanism. The components are fabricated with ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120. | |||
===Applicable Code Edition and Addenda=== | === | ||
The ASME Code, Section XI, 2004 Edition, No Addenda. | Applicable Code Edition and Addenda=== | ||
The ASME Code, Section XI, 2004 Edition, No Addenda. | |||
===Applicable Code Requirement=== | === | ||
Applicable Code Requirement=== | |||
The examination requirements for Class 1 piping welds are governed by the risk-informed inservice inspection program that was approved by the NRC in a {{letter dated|date=March 27, 2012|text=letter dated March 27, 2012}}, (ADAMS Accession No. ML120800433). The program was developed in accordance with ASME Code Case N-716, Alternative Piping Classification and Examination Requirements. | The examination requirements for Class 1 piping welds are governed by the risk-informed inservice inspection program that was approved by the NRC in a {{letter dated|date=March 27, 2012|text=letter dated March 27, 2012}}, (ADAMS Accession No. ML120800433). The program was developed in accordance with ASME Code Case N-716, Alternative Piping Classification and Examination Requirements. | ||
ASME Code Case N-716, Table 1, Examination Category R-A, requires that essentially 100 percent of the weld volume be examined. The alternative requirements of ASME Code Case N-460, approved for use in Regulatory Guide 1.147, Revision 18, allows credit for essentially 100 percent coverage provided greater than 90 percent of the required volume has been examined. | ASME Code Case N-716, Table 1, Examination Category R-A, requires that essentially 100 percent of the weld volume be examined. The alternative requirements of ASME Code Case N-460, approved for use in Regulatory Guide 1.147, Revision 18, allows credit for essentially 100 percent coverage provided greater than 90 percent of the required volume has been examined. | ||
The regulation at 10 CFR 50.55a(b)(2)(xv)(A), requires the following examination coverage criteria when applying Supplement 2 (Qualification Requirements for Wrought Austenitic Piping Welds) to Section XI Mandatory Appendix VIII (Performance Demonstration for Ultrasonic Examination Systems): | The regulation at 10 CFR 50.55a(b)(2)(xv)(A), requires the following examination coverage criteria when applying Supplement 2 (Qualification Requirements for Wrought Austenitic Piping Welds) to Section XI Mandatory Appendix VIII (Performance Demonstration for Ultrasonic Examination Systems): | ||
| Line 81: | Line 76: | ||
There are currently no PDI-qualified single-side examination procedures that demonstrate equivalency to two-sided examination procedures on austenitic piping welds. Current technology is not capable of reliably detecting or sizing flaws on the far side of an austenitic weld for configurations common to domestic nuclear applications. | There are currently no PDI-qualified single-side examination procedures that demonstrate equivalency to two-sided examination procedures on austenitic piping welds. Current technology is not capable of reliably detecting or sizing flaws on the far side of an austenitic weld for configurations common to domestic nuclear applications. | ||
The licensee stated that the PDI Performance Demonstration Qualification Summary certificates for austenitic piping list the limitation that single-side examination is performed on a best effort basis. The best effort qualification is provided in place of a complete single-side qualification to demonstrate that the examiners qualification and the subsequent weld examination is based on application of the best available technology. | The licensee stated that the PDI Performance Demonstration Qualification Summary certificates for austenitic piping list the limitation that single-side examination is performed on a best effort basis. The best effort qualification is provided in place of a complete single-side qualification to demonstrate that the examiners qualification and the subsequent weld examination is based on application of the best available technology. | ||
The licensee stated that the ASME Code required volume of these welds was interrogated ultrasonically to the maximum extent possible. No alternative methods or advanced | The licensee stated that the ASME Code required volume of these welds was interrogated ultrasonically to the maximum extent possible. No alternative methods or advanced technologies were considered capable of obtaining complete coverage of the examination volume. | ||
Table 1: Examination Category R-A Welds with Limited Volumetric Coverage Weld Identification Code Item No. | |||
technologies were considered capable of obtaining complete coverage of the examination volume. | System/Code Class Configuration Material Examination Angle and Wave Mode Limitations and Results Examination Coverage BPV-C-5049B R1.20 Reactor Coolant System/Class 1 Pipe to Flange ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120. | ||
Table 1: Examination Category R-A Welds with Limited Volumetric Coverage Code | 45° Shear Wave 60° Shear Wave 70° Shear Wave Single sided examination was performed due to pipe-to-flange configuration. | ||
System/Class 1 | No recordable indications were detected. | ||
Stainless Steel, Line | 56.25% | ||
BPV-C-5051B R1.20 Reactor Coolant System/Class 1 Pipe to Flange ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120. | |||
BPV-C-5051B | 45° Shear Wave 60° Shear Wave 70° Shear Wave Single sided examination was performed due to pipe-to-flange configuration. No recordable indications were detected. | ||
System/Class 1 | 54.13% | ||
Burden Caused by Compliance The licensee stated that a significant design modification or replacement of components with a different design that would allow full examination from both sides of the welds. The licensee stated this would be impractical based on cost, increased radiation exposure, and impact to plant equipment. | Burden Caused by Compliance The licensee stated that a significant design modification or replacement of components with a different design that would allow full examination from both sides of the welds. The licensee stated this would be impractical based on cost, increased radiation exposure, and impact to plant equipment. | ||
Proposed Alternative and Basis for Use The licensee stated that based on the volumetric coverage that was obtained with acceptable results and the visual (VT-2) examination performed during each refueling outage, it is reasonable to conclude that service-induced degradation would be detected. Therefore, these proposed alternatives will provide an acceptable level of quality and safety by providing reasonable assurance of structural integrity of the subject welds. | Proposed Alternative and Basis for Use The licensee stated that based on the volumetric coverage that was obtained with acceptable results and the visual (VT-2) examination performed during each refueling outage, it is reasonable to conclude that service-induced degradation would be detected. Therefore, these proposed alternatives will provide an acceptable level of quality and safety by providing reasonable assurance of structural integrity of the subject welds. | ||
Duration of Proposed Alternative The licensee requested relief for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | Duration of Proposed Alternative The licensee requested relief for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | ||
===NRC Staff Evaluation=== | ===NRC Staff Evaluation=== | ||
Applicable to all the subject connections, ASME Code Case N-716, Table 3, Examination Category R-A requires essentially 100 percent volumetric and surface examinations. However, complete volumetric examinations are restricted by materials and geometric configurations. To gain access for examination, the welds would require design modifications or component replacements with a design change that would allow an examination from both sides. The modification or replacement would increase cost, increase radiation exposure and impact plant equipment. The NRC staff determined that this would place an undue burden on the licensee, | Applicable to all the subject connections, ASME Code Case N-716, Table 3, Examination Category R-A requires essentially 100 percent volumetric and surface examinations. However, complete volumetric examinations are restricted by materials and geometric configurations. To gain access for examination, the welds would require design modifications or component replacements with a design change that would allow an examination from both sides. The modification or replacement would increase cost, increase radiation exposure and impact plant equipment. The NRC staff determined that this would place an undue burden on the licensee, thus, obtaining 100 percent of ASME Code required volumetric examinations is considered impractical. | ||
The subject welds received a volumetric examination to the maximum extent practical utilizing the best available techniques, which were qualified through the PDI for Supplement 2, with demonstrated best effort for single-sided examination from the accessible side of the weld. The subject welds were examined with a manual ultrasonic technique utilizing personnel, equipment and procedures qualified in accordance with ASME Code, Section XI, Appendix VIII, as implemented by the PDI. No recordable indications were detected. The components are monitored for through-wall leakage as part of the ASME Code, Section XI, System Pressure Test Program and receive a VT-2 examination at the end of each refueling outage during the system leakage tests as required by Section XI, Table IWB-2500-1, Category B-P for Class 1 components. Based on the above, the staff determined that the examinations obtained provide reasonable assurance of structural integrity of the subject welds. | |||
thus, obtaining 100 percent of ASME Code required volumetric examinations is considered impractical. | |||
The subject welds received a volumetric examination to the maximum extent practical utilizing the best available techniques, which were qualified through the PDI for Supplement 2, with demonstrated best effort for single-sided examination from the accessible side of the weld. The subject welds were examined with a manual ultrasonic technique utilizing personnel, equipment and procedures qualified in accordance with ASME Code, Section XI, Appendix VIII, as implemented by the PDI. No recordable indications were detected. The components are monitored for through-wall leakage as part of the ASME Code, Section XI, System Pressure Test Program and receive a VT-2 examination at the end of each refueling outage during the system leakage tests as required by Section XI, Table IWB-2500-1, Category B-P for Class 1 components. Based on the above, the staff determined that the examinations obtained provide reasonable assurance of structural integrity of the subject welds. | |||
==4.0 CONCLUSION== | |||
As discussed above, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Accordingly, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, the NRC staff grants the use of RR-04-32 for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | As discussed above, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Accordingly, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, the NRC staff grants the use of RR-04-32 for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020. | ||
All other ASME Code, Section XI, requirements for which relief has not been specifically requested and approved in this RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. | All other ASME Code, Section XI, requirements for which relief has not been specifically requested and approved in this RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. | ||
Principal Contributor: | Principal Contributor: | ||
E. Reichelt Date: August 2, 2021 | |||
ML21174A020 OFFICE | ML21174A020 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LAiT NRR/DORL/LPL3/LA NAME RGuzman KEntz SRohrer DATE 7/21/2021 7/21/2021 7/21/2021 OFFICE NRR/DNRL/NPHP/BC NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME MMitchell JDanna RGuzman DATE 4/21/2021 7/30/2021 8/2/2021}} | ||
Latest revision as of 08:23, 29 November 2024
| ML21174A020 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/02/2021 |
| From: | James Danna NRC/NRR/DORL/LPL1 |
| To: | Stoddard D Dominion Energy Nuclear Connecticut |
| Guzman R | |
| References | |
| EPID L-2020-LLR-0130 | |
| Download: ML21174A020 (8) | |
Text
August 2, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 2 - RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATIONS PERFORMED IN THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (EPID L-2020-LLR-0130)
Dear Mr. Stoddard:
By letter dated September 23, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20268A225), as supplemented by letter dated January 14, 2021 (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc. (the licensee) submitted relief requests (RRs) RR-04-31 and RR-04-32, which requested relief from the volumetric examination coverage requirements, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(5)(iii) on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry and materials of construction of the subject components for the Millstone Power Station, Unit No. 2 (Millstone 2). The relief is applicable to the third period of the fourth 10-year inservice inspection (ISI) interval for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020.
In its letter dated January 14, 2021, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage.
The U.S. Nuclear Regulatory Commission (NRC or Commission) staff has completed its review of the licensee's RR-04-32 request for Millstone 2. Pursuant to 10 CFR 50.55a(g)(6)(i), the NRC staff has determined that it is impractical for the licensee to comply with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, requirement; that the examinations performed to the extent practical provide reasonable assurance of structural integrity and leak tightness of the subject welds; and that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the staff grants relief for
D. Stoddard the subject examinations of the components contained in RR-04-32 for the fourth 10-year ISI interval at Millstone 2.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
If you have any questions, please contact the Millstone project manager, Richard Guzman, at 301-415-1030 or by e-mail to Richard.Guzman@nrc.gov.
Sincerely, James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosure:
Safety Evaluation cc: Listserv James G.
Danna Digitally signed by James G. Danna Date: 2021.08.02 11:19:18 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR LIMITED COVERAGE EXAMINATION PERFORMED IN THE FOURTH 10-YEAR INSPECTION INTERVAL DOMINION ENERGY NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336
1.0 INTRODUCTION
By letter dated September 23, 2020 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML20268A225), as supplemented by letter dated January 14, 2021 (ADAMS Accession No. ML21014A427), Dominion Energy Nuclear Connecticut, Inc., (the licensee) submitted Relief Request (RR) RR-04-32 from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2004 Edition, under the provisions of Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.55a(g)(5)(iii), for limited coverage examinations performed in the third inspection period of the fourth 10-year inservice inspection (ISI) interval for Millstone Power Station, Unit No. 2 (Millstone 2). The fourth 10-year ISI interval began on April 1, 2010, and ended on March 31, 2020.
In its letter dated January 14, 2021, the licensee withdrew RR-04-31 following an identified correction to the calculated examination coverage for the subject welds, resulting in greater than 90 percent examination coverage.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief on the basis that the required examination coverage was impractical due to physical obstructions and limitations imposed by design, geometry, and materials of construction, of the subject components.
2.0 REGULATORY EVALUATION
Regulation 10 CFR 50.55a(g) requires that the ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, and applicable editions and addenda.
Regulation 10 CFR 50.55a(g)(4) mandates the use of the ASME Code,Section XI, which states, in part, that ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a) twelve months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b).
Regulation 10 CFR 50.55a(g)(5)(iii), requires that if a licensee has determined that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit information to support the determinations.
Regulation 10 CFR 50.55a(g)(5)(iv), requires that where an examination requirement by the Code or Addenda is determined to be impractical by a licensee and is not included in the revised ISI program as permitted by paragraph 10 CFR 50.55a(g)(4), the basis for this determination must be demonstrated to the satisfaction of the Commission not later than 12 months after the expiration of the initial 120 month period of operation from the start of facility commercial operation and each subsequent 120 month period of operation during which the examination is determined to be impractical.
Regulation 10 CFR 50.55a(g)(6)(i) states in part that the Commission will evaluate determinations, under paragraph 10 CFR 50.55a(g)(5), that Code requirements are impractical.
The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request the use of an alternative and the NRC to grant the relief requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Relief Request RR-04-32 ASME Code Components Affected The affected components are ASME Code Class 1 welds, BPV-C-5049B and BPV-C-5051B, as shown in the table below. In accordance with Table IWB-2500-1 of the ASME Code,Section XI, the subject welds are classified as Examination Category R-A, Risk-Informed Piping Examinations, Item No. R1.20, Elements not Subject to a Degradation Mechanism. The components are fabricated with ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120.
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Applicable Code Edition and Addenda===
The ASME Code,Section XI, 2004 Edition, No Addenda.
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Applicable Code Requirement===
The examination requirements for Class 1 piping welds are governed by the risk-informed inservice inspection program that was approved by the NRC in a letter dated March 27, 2012, (ADAMS Accession No. ML120800433). The program was developed in accordance with ASME Code Case N-716, Alternative Piping Classification and Examination Requirements.
ASME Code Case N-716, Table 1, Examination Category R-A, requires that essentially 100 percent of the weld volume be examined. The alternative requirements of ASME Code Case N-460, approved for use in Regulatory Guide 1.147, Revision 18, allows credit for essentially 100 percent coverage provided greater than 90 percent of the required volume has been examined.
The regulation at 10 CFR 50.55a(b)(2)(xv)(A), requires the following examination coverage criteria when applying Supplement 2 (Qualification Requirements for Wrought Austenitic Piping Welds) to Section XI Mandatory Appendix VIII (Performance Demonstration for Ultrasonic Examination Systems):
(1) Piping must be examined in two axial directions and when examination in the circumferential direction is required, the circumferential examination must be performed in two directions, provided access is available.
(2) Where examination from both sides is not possible, full coverage credit may be claimed from a single side for ferritic welds. Where examination from both sides is not possible on austenitic welds, full coverage credit from a single side may be claimed only after completing a successful single sided Appendix VIII demonstration using flaws on the opposite side of the weld.
The regulation at 10 CFR 50.55a(b)(2)(xvi)(B), requires that examination performed from one side of a ferritic or stainless steel pipe weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two-sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as modified by 10 CFR 50.55a(b)(2)(xvi)(B) and 10 CFR 50.55a(b)(2)(xv)(A).
Impracticality of Compliance The licensee requested relief from the essentially 100 percent volumetric examination coverage requirement of the ASME Code,Section XI, for the subject austenitic welds due to the geometric configuration which limits the volumetric coverage that can be obtained.
The licensee stated that the subject welds were examined with a manual ultrasonic technique utilizing personnel, equipment and procedures qualified in accordance with ASME Code,Section XI, Appendix VIII, as implemented by the Performance Demonstration Initiative (PDI).
There are currently no PDI-qualified single-side examination procedures that demonstrate equivalency to two-sided examination procedures on austenitic piping welds. Current technology is not capable of reliably detecting or sizing flaws on the far side of an austenitic weld for configurations common to domestic nuclear applications.
The licensee stated that the PDI Performance Demonstration Qualification Summary certificates for austenitic piping list the limitation that single-side examination is performed on a best effort basis. The best effort qualification is provided in place of a complete single-side qualification to demonstrate that the examiners qualification and the subsequent weld examination is based on application of the best available technology.
The licensee stated that the ASME Code required volume of these welds was interrogated ultrasonically to the maximum extent possible. No alternative methods or advanced technologies were considered capable of obtaining complete coverage of the examination volume.
Table 1: Examination Category R-A Welds with Limited Volumetric Coverage Weld Identification Code Item No.
System/Code Class Configuration Material Examination Angle and Wave Mode Limitations and Results Examination Coverage BPV-C-5049B R1.20 Reactor Coolant System/Class 1 Pipe to Flange ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120.
45° Shear Wave 60° Shear Wave 70° Shear Wave Single sided examination was performed due to pipe-to-flange configuration.
No recordable indications were detected.
56.25%
BPV-C-5051B R1.20 Reactor Coolant System/Class 1 Pipe to Flange ASTM A-376, Type 316 Stainless Steel, Line Class BCA, Schedule 120.
45° Shear Wave 60° Shear Wave 70° Shear Wave Single sided examination was performed due to pipe-to-flange configuration. No recordable indications were detected.
54.13%
Burden Caused by Compliance The licensee stated that a significant design modification or replacement of components with a different design that would allow full examination from both sides of the welds. The licensee stated this would be impractical based on cost, increased radiation exposure, and impact to plant equipment.
Proposed Alternative and Basis for Use The licensee stated that based on the volumetric coverage that was obtained with acceptable results and the visual (VT-2) examination performed during each refueling outage, it is reasonable to conclude that service-induced degradation would be detected. Therefore, these proposed alternatives will provide an acceptable level of quality and safety by providing reasonable assurance of structural integrity of the subject welds.
Duration of Proposed Alternative The licensee requested relief for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020.
NRC Staff Evaluation
Applicable to all the subject connections, ASME Code Case N-716, Table 3, Examination Category R-A requires essentially 100 percent volumetric and surface examinations. However, complete volumetric examinations are restricted by materials and geometric configurations. To gain access for examination, the welds would require design modifications or component replacements with a design change that would allow an examination from both sides. The modification or replacement would increase cost, increase radiation exposure and impact plant equipment. The NRC staff determined that this would place an undue burden on the licensee, thus, obtaining 100 percent of ASME Code required volumetric examinations is considered impractical.
The subject welds received a volumetric examination to the maximum extent practical utilizing the best available techniques, which were qualified through the PDI for Supplement 2, with demonstrated best effort for single-sided examination from the accessible side of the weld. The subject welds were examined with a manual ultrasonic technique utilizing personnel, equipment and procedures qualified in accordance with ASME Code,Section XI, Appendix VIII, as implemented by the PDI. No recordable indications were detected. The components are monitored for through-wall leakage as part of the ASME Code,Section XI, System Pressure Test Program and receive a VT-2 examination at the end of each refueling outage during the system leakage tests as required by Section XI, Table IWB-2500-1, Category B-P for Class 1 components. Based on the above, the staff determined that the examinations obtained provide reasonable assurance of structural integrity of the subject welds.
4.0 CONCLUSION
As discussed above, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Accordingly, the NRC staff determines that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, the NRC staff grants the use of RR-04-32 for the fourth 10-year ISI for Millstone 2, which began on April 1, 2010, and ended on March 31, 2020.
All other ASME Code,Section XI, requirements for which relief has not been specifically requested and approved in this RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor:
E. Reichelt Date: August 2, 2021
ML21174A020 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LAiT NRR/DORL/LPL3/LA NAME RGuzman KEntz SRohrer DATE 7/21/2021 7/21/2021 7/21/2021 OFFICE NRR/DNRL/NPHP/BC NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME MMitchell JDanna RGuzman DATE 4/21/2021 7/30/2021 8/2/2021