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| {{#Wiki_filter:May 1, 2007Mr. David A. ChristianSr. Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. | | {{#Wiki_filter:May 1, 2007 Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc. |
| Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 | | Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 - RELIEF REQUESTSNOS. RR-89-60 AND IR-2-44 REQUESTING FOR ALTERNATIVES TO THE 10-YEAR REACTOR VESSEL EXAMINATIONS REQUIREMENTS (TAC NOS. MD1719 AND MD1720) | | MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 RELIEF REQUESTS NOS. RR-89-60 AND IR-2-44 REQUESTING FOR ALTERNATIVES TO THE 10-YEAR REACTOR VESSEL EXAMINATIONS REQUIREMENTS (TAC NOS. MD1719 AND MD1720) |
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| ==Dear Mr. Christian:== | | ==Dear Mr. Christian:== |
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| By letter dated May 11, 2006, [Agency Documents Access and Management System (ADAMS)accession number ML061530137] Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U.S. Nuclear Regulatory Commission (NRC), Relief Requests RR-89-60 and IR-2-44 for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI at the Millstone Power Station, Unit Nos. 2 and 3. Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code, Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. Therelief is requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR),Section 50.55a(a)(3)(i).Based upon the review of the information you provided, the NRC concluded that the proposedalternative provides reasonable assurance of satisfactory ultrasonic testing examination of thereactor pressure vessel shell-to-flange weld that are performed from the vessel side of the weld, and the NRC finds that the use PDI qualified procedures as an alternative provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), your proposed alternative is authorized. The NRC staff's Safety Evaluation is enclosed. If you have any questions, please contact the project manager, John Hughey at (301) 415-3204.Sincerely,/ra/Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-336, 50-423 | | By letter dated May 11, 2006, [Agency Documents Access and Management System (ADAMS) accession number ML061530137] Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U.S. Nuclear Regulatory Commission (NRC), Relief Requests RR-89-60 and IR-2-44 for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI at the Millstone Power Station, Unit Nos. 2 and 3. Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code, Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. The relief is requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR), |
| | Section 50.55a(a)(3)(i). |
| | Based upon the review of the information you provided, the NRC concluded that the proposed alternative provides reasonable assurance of satisfactory ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld that are performed from the vessel side of the weld, and the NRC finds that the use PDI qualified procedures as an alternative provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), your proposed alternative is authorized. The NRC staffs Safety Evaluation is enclosed. If you have any questions, please contact the project manager, John Hughey at (301) 415-3204. |
| | Sincerely, |
| | /ra/ |
| | Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-336, 50-423 |
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| ==Enclosure:== | | ==Enclosure:== |
| As stated cc: See next page | | As stated cc: See next page |
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| ML070740465*By memo datedNRR-028OFFICELPL1-2/PMLPL1-2/LALPL1-2/PMCPNB/BCOGC (NLO)LPLI-2/BCNAMEVNersesCSolaJHugheyTChan*DRothHChernoffDATE3/22/073/20/074/27/07 2/14/0704/25/075/1/07 Millstone Power Station, Unit Nos. 2 and 3 cc: | | ML070740465 *By memo dated NRR-028 OFFICE LPL1-2/PM LPL1-2/LA LPL1-2/PM CPNB/BC OGC (NLO) LPLI-2/BC NAME VNerses CSola JHughey TChan* DRoth HChernoff DATE 3/22/07 3/20/07 4/27/07 2/14/07 04/25/07 5/1/07 |
| Lillian M. Cuoco, EsquireSenior Counsel Dominion Resources Services, Inc.
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| Building 475, 5 th Floor Rope Ferry Road Waterford, CT 06385Edward L. Wilds, Jr., Ph.D.Director, Division of Radiation Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127Regional Administrator, Region IU.S. Nuclear Regulatory Commission
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| 475 Allendale Road King of Prussia, PA 19406First SelectmenTown of Waterford 15 Rope Ferry Road Waterford, CT 06385Charles Brinkman, DirectorWashington Operations Nuclear Services Westinghouse Electric Company 12300 Twinbrook Pkwy, Suite 330 Rockville, MD 20852Senior Resident InspectorMillstone Power Station c/o U.S. Nuclear Regulatory Commission P. O. Box 513 Niantic, CT 06357Mr. J. W. "Bill" Sheehan Co-Chair NEAC 19 Laurel Crest Drive Waterford, CT 06385Ms. Nancy Burton147 Cross Highway Redding Ridge, CT 00870Mr. Evan W. WoollacottCo-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road Simsbury, CT 06070Mr. Joseph RoyDirector of Operations Massachusetts Municipal Wholesale Electric Company P.O. Box 426 Ludlow, MA 01056Mr. David W. DodsonLicensing Supervisor Dominion Nuclear Connecticut, Inc.
| | Millstone Power Station, Unit Nos. 2 and 3 cc: |
| Building 475, 5 th FloorRoper Ferry Road Waterford, CT 06385Mr. J. Alan Price Site Vice President Dominion Nuclear Connecticut, Inc. | | Mr. Evan W. Woollacott Lillian M. Cuoco, Esquire Co-Chair Senior Counsel Nuclear Energy Advisory Council Dominion Resources Services, Inc. 128 Terrys Plain Road Building 475, 5th Floor Simsbury, CT 06070 Rope Ferry Road Waterford, CT 06385 Mr. Joseph Roy Director of Operations Edward L. Wilds, Jr., Ph.D. Massachusetts Municipal Wholesale Director, Division of Radiation Electric Company Department of Environmental P.O. Box 426 Protection Ludlow, MA 01056 79 Elm Street Hartford, CT 06106-5127 Mr. David W. Dodson Licensing Supervisor Regional Administrator, Region I Dominion Nuclear Connecticut, Inc. |
| Building 475, 5 th FloorRope Ferry Road Waterford, CT 06385 Mr. Chris L. FunderburkDirector, Nuclear Licensing and Operations Support Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 EnclosureSAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELIEF REQUESTS NOS. RR-89-60 AND IR-2-44MILLSTONE POWER STATION, UNIT NOS. 2 AND 3DOMINION NUCLEAR CONNECTICUT, INC.DOCKET NUMBERS 50-336 AND 50-42 | | U.S. Nuclear Regulatory Commission Building 475, 5th Floor 475 Allendale Road Roper Ferry Road King of Prussia, PA 19406 Waterford, CT 06385 First Selectmen Mr. J. Alan Price Town of Waterford Site Vice President 15 Rope Ferry Road Dominion Nuclear Connecticut, Inc. |
| | Waterford, CT 06385 Building 475, 5th Floor Rope Ferry Road Charles Brinkman, Director Waterford, CT 06385 Washington Operations Nuclear Services Westinghouse Electric Company Mr. Chris L. Funderburk 12300 Twinbrook Pkwy, Suite 330 Director, Nuclear Licensing and Rockville, MD 20852 Operations Support Innsbrook Technical Center Senior Resident Inspector 5000 Dominion Boulevard Millstone Power Station Glen Allen, VA 23060-6711 c/o U.S. Nuclear Regulatory Commission P. O. Box 513 Niantic, CT 06357 Mr. J. W. "Bill" Sheehan Co-Chair NEAC 19 Laurel Crest Drive Waterford, CT 06385 Ms. Nancy Burton 147 Cross Highway Redding Ridge, CT 00870 |
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| ==31.0 INTRODUCTION==
| | SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS NOS. RR-89-60 AND IR-2-44 MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 DOMINION NUCLEAR CONNECTICUT, INC. |
| By letter dated May 11, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted to theU.S. Nuclear Regulatory Commission (NRC), a request for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI at the Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3). Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code, Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. The relief isrequested pursuant to Title 10 of the Code of Federal Regulations (10 CFR),Section 50.55a(a)(3)(i).
| | DOCKET NUMBERS 50-336 AND 50-423 |
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| ==2.0REGULATORY EVALUATION== | | ==1.0 INTRODUCTION== |
| The inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is to beperformed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by NRC pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (includingsupports) shall 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 ISI 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(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ISI Code of record for the third 10-year ISI interval for MPS2and the second 10-year ISI interval for MPS3 is the 1989 Edition of the ASME Code, Section XI. In addition, 10 CFR 50.55a(g)(6)(ii)(C)(2) requires that licensees using the 1989 Edition or earlier editions must implement the 1995 Edition with the 1996 Addenda of Appendix VIII and the supplements to Appendix VIII of Section XI of the ASME Code.3.0EVALUATION FOR RELIEF REQUEST NOS. RR-89-60 AND IR-2-443.1Components for Which Relief is RequestedASME Code Class 1, Item B1.30, Reactor Pressure Vessel (RPV) Shell-to-Flange weldsidentified as:Millstone 2:Weld FS-1Millstone 3:Weld 101-1213.2Code RequirementsThe 1989 Edition of ASME Code, Section XI, Appendix I, Subparagraph I-2110, requires thatultrasonic testing (UT) of RPV shell-to-flange welds be conducted in accordance with Article 4 of ASME Code, Section V, supplemented by the requirements of Table I-2000-1. In addition, Regulatory Guide (RG) 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," serves as guidance for UT examination of RPV welds.3.3 Licensee's Proposed Alternative and Basis for UseDuring the upcoming 10-year RPV vessel examinations, the licensee proposes to performultrasonic examinations of the RPV shell-to-flange welds using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR50.55a. The examination will be performed automated as qualified in accordance with ASME Code, Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplement s 4 and 6 asamended by 10 CFR 50.55a and the PDI Program. Since the subject examinations will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single side access for examination of these welds.Appendix VIII requirements were developed and adopted to ensure the effectiveness ofultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations. The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to or surpass the requirements of the ASME Code, Section V, Article 4 ultrasonic examination requirements. The licensee finds that though Appendix VIII is not required for the RPV shell-to-flange weldexamination, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing offlaws in the subject welds will be equal to or exceed the requirements of ASME Code, Section V, Article 4 and the guidance in RG 1.150. 3.4NRC Staff Evaluation The ASME Code requires that ultrasonic examination of shell-to-flange welds in vessels greaterthan 2 inches in thickness be conducted in accordance with Article 4 of the ASME Code, Section V, as supplemented by requirements in Table I-2000-1. ASME Code, Section V, Article 4 provides a prescriptive process for qualifying UT procedures and performing examinations. The licensee instead proposes to use procedures and personnel qualified in accordance with performance-based criteria listed in the 1995 Edition, 1996 Addenda of the ASME Code, Section XI, Appendix VIII, Supplements 4 and 6 as implemented by the industry'sPDI program. These performance-based methods are currently required by 10 CFR 50.55a for examination of all other RPV shell welds (having replaced the Article 4 techniques).Amplitude-based sizing techniques such as the prescriptive UT procedures that comply with therequirements of Article 4 of ASME Code, Section V, are based on the amplitude of the returned signal and correlating that amplitude with an equivalent machined reflector such as a notch or a side-drilled hole. However, correlation between defect size and amplitude has been poor.
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| This is not unexpected given the number of variables from the material, equipment and defect itself. The material has potential velocity and microstructure variations, and the equipment has potential amplitude variations due to the type of pulser, frequency band, cabling, and other inherent electrical parameters. Perhaps the biggest variable is the defect itself. Ultrasonic examination is highly sensitive to defect orientation. Also, transparency, roughness, curvature, and location play a role in the ability to detect and size defects. In addition, conventional amplitude-based ultrasonics is particularly unreliable for vertical defects.When prescriptive UT procedures that comply with the requirements of Article 4 of ASME Code,Section V, were used in round robin tests containing real flaws in RPV mockups, and the results statistically analyzed according to the screening criteria of ASME Code, Section XI, Appendix VIII, the procedures proved to be less effective than ex aminations that utilizeAppendix VIII, Supplements 4 and 6, qualified procedures. Performance-based UT is generallyapplied with higher sensitivity, which increases the probability of detecting a flaw when compared to prescriptive Section V, Article 4 requirements. Also, flaw sizing is more accurately determined with the time-based tip diffraction criteria used by performance-based ultrasonics than with the less accurate amplitude criteria for prescriptive Section V, Article 4 requirement.
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| Procedures, equipment, and personnel qualified through the PDI program have demonstrated their skill level to detect flaws common to nuclear power plants and have shown high probability of detection levels. This has resulted in an increased reliability of inspections for weldconfigurations subject to the requirements of Appendix VIII.The licensee states that the examination, using automated equipment, will be performed byprocedures qualified in accordance with ASME Code, Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDIprogram. Due to the weld configuration the examination will be performed from a single side.
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| Therefore, all procedures, personnel, and equipment will be qualified for single-sided access for examination of the subject welds. 3.5SummaryThe NRC staff finds that the use of UT procedures and personnel qualified to the 1995 Editionwith 1996 Addenda of Section XI of the ASME Code, Appendix VIII, Supplement 4 and 6, asmodified by 10 CFR 50.55a by demonstration through the PDI program for the RPV shell-to-flange weld, provides equivalent or better examination results than those obtained from ASME Code, Section V requirements and RG 1.150 recommendations. Therefore, based on the above analysis, the staff considers that an acceptable level of quality and safety will be maintained when using the licensee's proposed alternative examination.
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| ==4.0 CONCLUSION==
| | By letter dated May 11, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U.S. Nuclear Regulatory Commission (NRC), a request for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI at the Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3). Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code, Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. The relief is requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR), |
| | Section 50.55a(a)(3)(i). |
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| The NRC staff concludes that the proposed alternative with PDI qualified procedures andpersonnel applied from the RPV shell surface along with the improved capabilities as discussed above will provide equivalent or better examination results than those realized from the ASME Code, Section V, requirements and RG 1.150 recommendations and, therefore, will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC authorizes the proposed alternative for the remainder of the third 10-year ISI interval at MPS2 and the second 10-year ISI interval at MPS3.All other ASME Code, Section XI requirements for which relief was not specifically requestedand approved in this, or previous, relief requests remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.Principal Contributor: A. Keim Date: May 1, 2007}} | | ==2.0 REGULATORY EVALUATION== |
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| | The inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by NRC pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. |
| | Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall 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 ISI 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(b) twelve months prior to the start of the 120-month interval, subject to the limitations and Enclosure |
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| | modifications listed therein. The ISI Code of record for the third 10-year ISI interval for MPS2 and the second 10-year ISI interval for MPS3 is the 1989 Edition of the ASME Code, Section XI. In addition, 10 CFR 50.55a(g)(6)(ii)(C)(2) requires that licensees using the 1989 Edition or earlier editions must implement the 1995 Edition with the 1996 Addenda of Appendix VIII and the supplements to Appendix VIII of Section XI of the ASME Code. |
| | 3.0 EVALUATION FOR RELIEF REQUEST NOS. RR-89-60 AND IR-2-44 3.1 Components for Which Relief is Requested ASME Code Class 1, Item B1.30, Reactor Pressure Vessel (RPV) Shell-to-Flange welds identified as: |
| | Millstone 2: Weld FS-1 Millstone 3: Weld 101-121 3.2 Code Requirements The 1989 Edition of ASME Code, Section XI, Appendix I, Subparagraph I-2110, requires that ultrasonic testing (UT) of RPV shell-to-flange welds be conducted in accordance with Article 4 of ASME Code, Section V, supplemented by the requirements of Table I-2000-1. In addition, Regulatory Guide (RG) 1.150, Revision 1, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, serves as guidance for UT examination of RPV welds. |
| | 3.3 Licensees Proposed Alternative and Basis for Use During the upcoming 10-year RPV vessel examinations, the licensee proposes to perform ultrasonic examinations of the RPV shell-to-flange welds using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a. The examination will be performed automated as qualified in accordance with ASME Code, Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI Program. Since the subject examinations will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single side access for examination of these welds. |
| | Appendix VIII requirements were developed and adopted to ensure the effectiveness of ultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations. The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to or surpass the requirements of the ASME Code, Section V, Article 4 ultrasonic examination requirements. |
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| | The licensee finds that though Appendix VIII is not required for the RPV shell-to-flange weld examination, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing of flaws in the subject welds will be equal to or exceed the requirements of ASME Code, Section V, Article 4 and the guidance in RG 1.150. |
| | 3.4 NRC Staff Evaluation The ASME Code requires that ultrasonic examination of shell-to-flange welds in vessels greater than 2 inches in thickness be conducted in accordance with Article 4 of the ASME Code, Section V, as supplemented by requirements in Table I-2000-1. ASME Code, Section V, Article 4 provides a prescriptive process for qualifying UT procedures and performing examinations. The licensee instead proposes to use procedures and personnel qualified in accordance with performance-based criteria listed in the 1995 Edition, 1996 Addenda of the ASME Code, Section XI, Appendix VIII, Supplements 4 and 6 as implemented by the industrys PDI program. These performance-based methods are currently required by 10 CFR 50.55a for examination of all other RPV shell welds (having replaced the Article 4 techniques). |
| | Amplitude-based sizing techniques such as the prescriptive UT procedures that comply with the requirements of Article 4 of ASME Code, Section V, are based on the amplitude of the returned signal and correlating that amplitude with an equivalent machined reflector such as a notch or a side-drilled hole. However, correlation between defect size and amplitude has been poor. |
| | This is not unexpected given the number of variables from the material, equipment and defect itself. The material has potential velocity and microstructure variations, and the equipment has potential amplitude variations due to the type of pulser, frequency band, cabling, and other inherent electrical parameters. Perhaps the biggest variable is the defect itself. Ultrasonic examination is highly sensitive to defect orientation. Also, transparency, roughness, curvature, and location play a role in the ability to detect and size defects. In addition, conventional amplitude-based ultrasonics is particularly unreliable for vertical defects. |
| | When prescriptive UT procedures that comply with the requirements of Article 4 of ASME Code, Section V, were used in round robin tests containing real flaws in RPV mockups, and the results statistically analyzed according to the screening criteria of ASME Code, Section XI, Appendix VIII, the procedures proved to be less effective than examinations that utilize Appendix VIII, Supplements 4 and 6, qualified procedures. Performance-based UT is generally applied with higher sensitivity, which increases the probability of detecting a flaw when compared to prescriptive Section V, Article 4 requirements. Also, flaw sizing is more accurately determined with the time-based tip diffraction criteria used by performance-based ultrasonics than with the less accurate amplitude criteria for prescriptive Section V, Article 4 requirement. |
| | Procedures, equipment, and personnel qualified through the PDI program have demonstrated their skill level to detect flaws common to nuclear power plants and have shown high probability of detection levels. This has resulted in an increased reliability of inspections for weld configurations subject to the requirements of Appendix VIII. |
| | The licensee states that the examination, using automated equipment, will be performed by procedures qualified in accordance with ASME Code, Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI program. Due to the weld configuration the examination will be performed from a single side. |
| | Therefore, all procedures, personnel, and equipment will be qualified for single-sided access for examination of the subject welds. |
| | |
| | 3.5 Summary The NRC staff finds that the use of UT procedures and personnel qualified to the 1995 Edition with 1996 Addenda of Section XI of the ASME Code, Appendix VIII, Supplement 4 and 6, as modified by 10 CFR 50.55a by demonstration through the PDI program for the RPV shell-to-flange weld, provides equivalent or better examination results than those obtained from ASME Code, Section V requirements and RG 1.150 recommendations. Therefore, based on the above analysis, the staff considers that an acceptable level of quality and safety will be maintained when using the licensees proposed alternative examination. |
| | |
| | ==4.0 CONCLUSION== |
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| | The NRC staff concludes that the proposed alternative with PDI qualified procedures and personnel applied from the RPV shell surface along with the improved capabilities as discussed above will provide equivalent or better examination results than those realized from the ASME Code, Section V, requirements and RG 1.150 recommendations and, therefore, will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC authorizes the proposed alternative for the remainder of the third 10-year ISI interval at MPS2 and the second 10-year ISI interval at MPS3. |
| | All other ASME Code, Section XI requirements for which relief was not specifically requested and approved in this, or previous, relief requests remain applicable, including third party review by the Authorized Nuclear Inservice Inspector. |
| | Principal Contributor: A. Keim Date: May 1, 2007}} |
Letter Sequence Other |
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MONTHYEARML0707404652007-05-0101 May 2007
Relief, Relief Requests Nos. RR-89-60 and IR-2-44 Requesting for Alternatives to the 10-Year Reactor Vessel Examinations Requirements
Project stage: Other
2007-05-01
[Table View] |
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Category:Code Relief or Alternative
MONTHYEARML23188A0432023-07-31031 July 2023
Authorization and Safety Evaluation for Alternative Request No. IR-04-11
ML22201A5082022-07-28028 July 2022
Authorization and Safety Evaluation for Alternative Request No. IR-04-09
ML21284A0062021-10-29029 October 2021
Authorization and Safety Evaluation for Alternative Request No. RR-05-04 and IR-4-02
ML21174A0202021-08-0202 August 2021
Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval
ML21167A3552021-07-16016 July 2021
Authorization and Safety Evaluation for Alternative Request No. RR-05-06
ML20312A0012020-12-10010 December 2020
Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval
ML20312A0022020-12-10010 December 2020
Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 Through L-2020-LLR-0032)
ML20287A4712020-10-20020 October 2020
Proposed Alternative RR-05-05 to the Requirements of the ASME Code Containment Unbonded Post-Tensioning System Inservice Inspection Requirements
ML20189A2062020-07-16016 July 2020
Relief Request IR-4-03 Concerning Non-Code Methodology to Demonstrate Structural Integrity of Class 3 Moderate-Energy Piping
ML20080K5082020-03-24024 March 2020
Alternative Request RR-05-03 for the Fifth 10-Year Inservice Inspection Interval
ML19340A0012019-12-18018 December 2019
Proposed Alternative Request IR-4-01 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
ML19338G3722019-12-18018 December 2019
Alternative Requests RR-05-01 and RR-05-02 for the Fifth 10-Year Inservice Inspection Interval
ML19340A0002019-12-13013 December 2019
Relief Request IR-3-39, Proposed Alternative to ASME Code Weld Preheat Requirements
ML19275D2522019-09-24024 September 2019
Alternative Request RR-05-03, Extension of ASME Code Case N-770-2 Volumetric Inspection Frequency for Reactor Coolant Pump Inlet and Outlet Nozzle Dissimilar Metal Butt Welds
ML19098A0342019-04-30030 April 2019
Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins
ML18290A6022018-11-13013 November 2018
Alternative Requests Related to the Fifth and Fourth 10-Year Interval Pump, Valve, and Snubber Inservice Testing Programs, Respectively (EPID L-2018-LLR- 0012 Through EPID L-2018-LLR-0022)
ML18252A0032018-09-17017 September 2018
Alternative Requests RR-04-27 and IR-3-38 for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
ML18066A5222018-02-28028 February 2018
Proposed Alternative Requests RR-04-27 and IR-3-38 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55(z)(1)
ML17132A1872017-05-25025 May 2017
Alternative Relief Requests RR-04-24 and IR-3-30: Reactor Pressure Vessel Threads in Flange
ML17135A2962017-05-25025 May 2017
Alternative Relief Request RR-04-25 Boric Acid Pump P-19B Stuffing Box Cover
ML17122A3742017-05-0303 May 2017
Alternative Relief Request RR-04-26 Boric Acid Pump P-19B Stuffing Box Cover
ML17125A2522017-04-28028 April 2017
ASME Section XI Relief Request RR-04-26
ML17090A1102017-03-29029 March 2017
ASME Section XI Relief Request RR-04-25
ML16363A0892017-01-23023 January 2017
Alternative Relief RR-04-23 and IR-3-28 from the Requirements of ASME Code Section XI Regarding Radiographic Examinations
ML16277A6782016-10-18018 October 2016
Alternative Request RR-04-22 to Implement Extended Reactor Vessel Inservice Inspection Interval
ML16172A1352016-07-13013 July 2016
Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval
ML16038A0012016-02-16016 February 2016
Alternative Request IR-3-27 for Implementation of Extended Reactor Vessel Inservice Inspection Interval
ML15257A0052015-09-21021 September 2015
Relief from the Requirements of ASME Code Section XI Regarding Radiographic Examinations
ML15216A3632015-07-30030 July 2015
ASME Section XI Inservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval
ML15216A3592015-07-30030 July 2015
ASME Section XI Inservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval
ML15082A4092015-04-24024 April 2015
Alternative Use of Weld Overlay as Repair and Mitigation Technique
ML14217A2032014-09-0404 September 2014
Relief from the Requirements of the ASME Code Section XI, Requirements for Repair/Replacement of Class 3 Service Water Valves (Tac No. MF1314)
ML14163A5862014-07-10010 July 2014
Relief from the Inservice Testing Requirements of American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants
ML14091A9732014-04-0404 April 2014
Issuance of Relief Request RR-04-16 Regarding Use of Encoded Phased Array Ultrasonic Examination in Lieu of Radiography
ML1130001002011-11-0909 November 2011
Issuance of Relief Request RR-04-12 Regarding the Temporary Non-Code Compliant Condition of the Class 3 Service Water System 10 Inch Emergency Diesel Generator Supply Piping Flange
ML11234A0772011-08-19019 August 2011
Relief Request RR-04-12 for the Temporary Non-Code Compliant Condition of the Class 3 Service Water System 10 Inch Emergency Diesel Generator Supply Piping Flange
ML1118810292011-07-27027 July 2011
Issuance of Relief Request RR-04-04 Regarding Use of Alternative Pressure Testing Requirements
ML1118706002011-07-22022 July 2011
Issuance of Relief Request RR-04-05 Regarding Use of Alternative Pressure Testing Requirements
ML1106800802011-03-24024 March 2011
Issuance of Relief Request lR-3-14 -- Use of Risk-Informed Inservice Inspection Program Plan
ML1014700992010-06-11011 June 2010
Issuance of Relief Request IR-3-05 Regarding Use of American Society of Mechanical Engineering Code, Section XI, Appendix Viii
ML1012412192010-05-13013 May 2010
Relief Request IR-3-02 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition
ML1010400422010-04-29029 April 2010
Issuance of Relief Request lR-3-11 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition
ML0935702372010-04-26026 April 2010
Issuance of Relief Request RR-89-67 Regarding the Repair of Reactor Coolant Pump Seal Cooler Return Tubing and Weld
ML1011301872010-04-19019 April 2010
ASME Section XI Inservice Inspection Program Relief Request for Limited Coverage Examinations Performed in the Second 10-Year Inspection Interval
ML1008407382010-04-15015 April 2010
Issuance of Relief Requests IR-3-13 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition
ML1006801182010-04-0606 April 2010
Issuance of Relief Request IR-3-01 Regarding Use of American Society of Mechanical Engineering Code, Section XI, Appendix Viii
ML1009002002010-03-30030 March 2010
Relief Requests RR-04-02, Alternative VT-2 Pressure Testing Requirements for the Lower Portion of the Reactor Pressure Vessel, and RR-04-03, Alternative Evaluation Criteria for Code Case N-513-2, Temporary Acceptance of Flaws In..
ML1006404462010-03-12012 March 2010
Issuance Relief
ML1005402202010-02-19019 February 2010
Relief Request IR-3-01 Supplemental Information Re Snubber Inspection and Testing for Third 10-Year Interval
ML0923901412009-08-24024 August 2009
Relief Request for Millstone Power Station, Unit 3, Relief Request IR-3-04, Response to Request for Additional Information for Alternative Brazed Joint Assessment Methodology
2023-07-31
[Table view]
Category:Letter
MONTHYEAR05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024
Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary
IR 05000336/20244022024-10-0808 October 2024
Security Baseline Inspection Report 05000336/2024402 and 05000423/2024402 (Cover Letter Only)
ML24281A1102024-10-0707 October 2024
Requalification Program Inspection
05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024
Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
ML24240A1692024-09-18018 September 2024
Cy 2023 Summary of Decommissioning Trust Fund Status
ML24260A2192024-09-16016 September 2024
Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification
ML24260A1952024-09-16016 September 2024
Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel
ML24248A2272024-09-0404 September 2024
Operator Licensing Examination Approval
ML24240A1532024-09-0303 September 2024
Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel
IR 05000336/20240052024-08-29029 August 2024
Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005
IR 05000336/20240022024-08-13013 August 2024
Integrated Inspection Report 05000336/2024002 and 05000423/2024002
ML24221A2872024-08-0808 August 2024
Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage
IR 05000336/20244412024-08-0606 August 2024
Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only)
ML24212A0742024-08-0505 August 2024
Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification
ML24211A1712024-07-25025 July 2024
Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change
IR 05000336/20244032024-07-22022 July 2024
Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403
IR 05000336/20245012024-07-0101 July 2024
Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501
ML24180A0932024-06-28028 June 2024
Readiness for Additional Inspection: EA-23-144
IR 05000336/20240102024-06-26026 June 2024
Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010
ML24178A2422024-06-25025 June 2024
2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46
IR 05000336/20244402024-06-24024 June 2024
Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only)
ML24281A2072024-06-20020 June 2024
Update to the Final Safety Analysis Report, Revision 37 (Redacted Version)
ML24177A2792024-06-20020 June 2024
Preparation and Scheduling of Operator Licensing Examinations
ML24280A0012024-06-20020 June 2024
Update to the Final Safety Analysis Report (Redacted Version)
ML24162A0882024-06-10010 June 2024
Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications
ML24170B0532024-06-10010 June 2024
DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code
ML24165A1292024-06-0505 June 2024
ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023
ML24128A2772024-06-0404 June 2024
Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary)
ML24151A6482024-06-0303 June 2024
Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities
ML24110A0562024-05-21021 May 2024
Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter)
ML24109A0032024-05-21021 May 2024
Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary)
ML24141A1502024-05-20020 May 2024
Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
ML24141A2432024-05-20020 May 2024
Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15
ML24142A0952024-05-20020 May 2024
End of Cycle 22 Steam Generator Tube Inspection Report
IR 05000336/20240012024-05-14014 May 2024
Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation
ML24123A2042024-05-0202 May 2024
Pre-Decisional Replay to EA-23-144
ML24123A2272024-05-0202 May 2024
Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
ML24123A1222024-04-30030 April 2024
Inservice Inspection Program - Owners Activity Report, Refueling Outage 22
IR 05000336/20244012024-04-30030 April 2024
Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only)
ML24116A0452024-04-25025 April 2024
Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only)
ML24114A2662024-04-24024 April 2024
Submittal of 2023 Annual Radioactive Effluent Release Report
ML24116A1742024-04-24024 April 2024
Annual Radiological Environmental Operating Report
ML24103A0202024-04-22022 April 2024
Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits
ML24106A2032024-04-15015 April 2024
2023 Annual Environmental Operating Report
ML24088A3302024-04-0404 April 2024
Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel
ML24093A2162024-04-0101 April 2024
Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits
ML24093A1022024-04-0101 April 2024
Alternative Request IR-4-13, Proposed Alternative Request to Support Steam Genera Tor Channel Head Drain Modification
IR 05000336/20240112024-04-0101 April 2024
Comprehensive Engineering Team Inspection - Inspection Report 05000336/2024011 and 05000423/2024011
ML24092A0752024-03-28028 March 2024
3R22 Refueling Outage Inservice Inspection (ISI) Owners Activity Report Extension
ML24088A2352024-03-26026 March 2024
Decommissioning Funding Status Report
2024-09-04
[Table view]
Category:Safety Evaluation
MONTHYEARML24128A2772024-06-0404 June 2024
Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary)
ML24109A0032024-05-21021 May 2024
Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary)
ML23341A0172024-01-12012 January 2024
Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures
ML23283A3052023-12-20020 December 2023
Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary)
ML23230A0502023-10-0202 October 2023
5 of the Quality Assurance Topical Report - Review of Program Changes
ML23226A0052023-09-26026 September 2023
Issuance of Amendment No. 287 Supplement to Spent Fuel Pool Criticality Safety Analysis
ML23188A0432023-07-31031 July 2023
Authorization and Safety Evaluation for Alternative Request No. IR-04-11
ML23175A0052023-07-12012 July 2023
Alternative Request P-07 for Pump Periodic Verification Testing Program for Containment Recirculation Spray System Pumps
ML23072A0892023-05-0101 May 2023
(Amendments 346 & 286), North Anna 1 & 2 (Amnds 294 & 277), Surry 1 & 2 (Amnds 311 & 311), and Summer 1 (Amd 225) - Issuance of Amendments to Revise TSs to Adopt TSTF-554 Revise Reactor Coolant Leakage Requirements
ML23058A4542023-03-16016 March 2023
Issuance of Amendment Nos. 345 and 285 Regarding Adoption of Technical Specification Task Force-359, Increase Flexibility in Mode Restraints
ML21320A0072022-09-0707 September 2022
Review of Appendix E to DOM-NAF-2, Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code (EPID L-2021-LLT-0000) (Non-Proprietary)
ML22201A5082022-07-28028 July 2022
Authorization and Safety Evaluation for Alternative Request No. IR-04-09
ML22095A1072022-07-11011 July 2022
Issuance of Amendment Nos. 120, 344, & 284, 293 & 276, & 307 & 307 to Relocate Requirements to the QAPD
ML22039A3392022-03-0303 March 2022
Request for Alternative Frequency to Supplemental Valve Position Verification Testing Requirements in the Fourth 10-year Valve Inservice Testing Program
ML22041A0102022-03-0101 March 2022
V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition
ML22007A1512022-02-16016 February 2022
Issuance of Amendment No. 282 Regarding Shutdown Bank Technical Specification Requirements and Alternate Control Rod Position Monitoring Requirements
ML21326A0992022-01-0707 January 2022
Issuance of Amendment No. 281 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-177642-P-A, Revision 1
ML21262A0012021-11-0909 November 2021
Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate
ML21284A0062021-10-29029 October 2021
Authorization and Safety Evaluation for Alternative Request No. RR-05-04 and IR-4-02
ML21227A0002021-10-0505 October 2021
Issuance of Amendment No. 279 Regarding Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss-of-Coolant Accident
ML21222A2302021-09-0909 September 2021
Issuance of Amendment No. 343 Revision to Technical Specifications for Steam Generator Inspection Frequency (L-2020-LLA-0227)
ML21174A0202021-08-0202 August 2021
Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval
ML21167A3552021-07-16016 July 2021
Authorization and Safety Evaluation for Alternative Request No. RR-05-06
ML21167A2112021-06-30030 June 2021
Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0081 Through L-2020-LLR-0088)
ML21075A0452021-03-26026 March 2021
Request to Utilize Code Case N-885
ML21043A1622021-03-25025 March 2021
Issuance of Amendment No. 278 Regarding Revision to Battery Surveillance Requirements
ML21026A1422021-02-23023 February 2021
Issuance of Amendment No. 342 Revision to Technical Specification Table 3.3-11, Accident Monitoring Instrumentation
ML20312A0022020-12-10010 December 2020
Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 Through L-2020-LLR-0032)
ML20312A0012020-12-10010 December 2020
Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval
ML20287A4712020-10-20020 October 2020
Proposed Alternative RR-05-05 to the Requirements of the ASME Code Containment Unbonded Post-Tensioning System Inservice Inspection Requirements
ML20275A0002020-10-14014 October 2020
Issuance of Amendment No. 277 to Revise Technical Specification 6.8.4.g to Allow a One-Time Deferral of the Steam Generator Inspections
ML20237H9952020-09-29029 September 2020
Issuance of Amendment No. 341 Revision to Technical Specification 6.25, Pre-Stressed Concrete Containment Tendon Surveillance Program
ML20252A0072020-09-15015 September 2020
Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI
ML20191A0042020-08-0707 August 2020
Issuance of Amendment No. 340 Revised Technical Specification Limits for Primary and Secondary Coolant Activity
ML20189A2062020-07-16016 July 2020
Relief Request IR-4-03 Concerning Non-Code Methodology to Demonstrate Structural Integrity of Class 3 Moderate-Energy Piping
ML20161A0002020-07-15015 July 2020
Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals
ML20140A3692020-06-24024 June 2020
Issuance of Amendment No. 339 Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time
ML20080K5082020-03-24024 March 2020
Alternative Request RR-05-03 for the Fifth 10-Year Inservice Inspection Interval
ML19340A0252020-01-30030 January 2020
Issuance of Amendment No. 337 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components of Nuclear Power Reactors
ML19305D2482019-12-31031 December 2019
Issuance of Amendments Adoption of Emergency Action Level Schemes Per NEI 99-01, Rev. 6
ML19340A0012019-12-18018 December 2019
Proposed Alternative Request IR-4-01 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
ML19338G3722019-12-18018 December 2019
Alternative Requests RR-05-01 and RR-05-02 for the Fifth 10-Year Inservice Inspection Interval
ML19340A0002019-12-13013 December 2019
Relief Request IR-3-39, Proposed Alternative to ASME Code Weld Preheat Requirements
ML19126A0002019-05-28028 May 2019
Issuance of Amendment No. 273 Regarding Technical Specification Changes for Spent Fuel Storage and New Fuel Storage
ML19098A0342019-04-30030 April 2019
Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins
ML19042A2772019-03-21021 March 2019
Issuance of Amendment No. 272 Regarding Revision to Technical Specification Action Statement for Loss of Control Building Inlet Ventilation Radiation Monitor Instrumentation Channels
ML18290A6022018-11-13013 November 2018
Alternative Requests Related to the Fifth and Fourth 10-Year Interval Pump, Valve, and Snubber Inservice Testing Programs, Respectively (EPID L-2018-LLR- 0012 Through EPID L-2018-LLR-0022)
ML18275A0122018-10-0404 October 2018
Alternative Request P-06 for the 'C' Charging Pump Test Frequency
ML18246A0072018-09-25025 September 2018
Issuance of Amendment No. 335 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals
ML18252A0032018-09-17017 September 2018
Alternative Requests RR-04-27 and IR-3-38 for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography
2024-06-04
[Table view] |
Text
May 1, 2007 Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 RELIEF REQUESTS NOS. RR-89-60 AND IR-2-44 REQUESTING FOR ALTERNATIVES TO THE 10-YEAR REACTOR VESSEL EXAMINATIONS REQUIREMENTS (TAC NOS. MD1719 AND MD1720)
Dear Mr. Christian:
By letter dated May 11, 2006, [Agency Documents Access and Management System (ADAMS) accession number ML061530137] Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U.S. Nuclear Regulatory Commission (NRC), Relief Requests RR-89-60 and IR-2-44 for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI at the Millstone Power Station, Unit Nos. 2 and 3. Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code,Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. The relief is requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a(a)(3)(i).
Based upon the review of the information you provided, the NRC concluded that the proposed alternative provides reasonable assurance of satisfactory ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld that are performed from the vessel side of the weld, and the NRC finds that the use PDI qualified procedures as an alternative provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), your proposed alternative is authorized. The NRC staffs Safety Evaluation is enclosed. If you have any questions, please contact the project manager, John Hughey at (301) 415-3204.
Sincerely,
/ra/
Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-336, 50-423
Enclosure:
As stated cc: See next page
ML070740465 *By memo dated NRR-028 OFFICE LPL1-2/PM LPL1-2/LA LPL1-2/PM CPNB/BC OGC (NLO) LPLI-2/BC NAME VNerses CSola JHughey TChan* DRoth HChernoff DATE 3/22/07 3/20/07 4/27/07 2/14/07 04/25/07 5/1/07
Millstone Power Station, Unit Nos. 2 and 3 cc:
Mr. Evan W. Woollacott Lillian M. Cuoco, Esquire Co-Chair Senior Counsel Nuclear Energy Advisory Council Dominion Resources Services, Inc. 128 Terrys Plain Road Building 475, 5th Floor Simsbury, CT 06070 Rope Ferry Road Waterford, CT 06385 Mr. Joseph Roy Director of Operations Edward L. Wilds, Jr., Ph.D. Massachusetts Municipal Wholesale Director, Division of Radiation Electric Company Department of Environmental P.O. Box 426 Protection Ludlow, MA 01056 79 Elm Street Hartford, CT 06106-5127 Mr. David W. Dodson Licensing Supervisor Regional Administrator, Region I Dominion Nuclear Connecticut, Inc.
U.S. Nuclear Regulatory Commission Building 475, 5th Floor 475 Allendale Road Roper Ferry Road King of Prussia, PA 19406 Waterford, CT 06385 First Selectmen Mr. J. Alan Price Town of Waterford Site Vice President 15 Rope Ferry Road Dominion Nuclear Connecticut, Inc.
Waterford, CT 06385 Building 475, 5th Floor Rope Ferry Road Charles Brinkman, Director Waterford, CT 06385 Washington Operations Nuclear Services Westinghouse Electric Company Mr. Chris L. Funderburk 12300 Twinbrook Pkwy, Suite 330 Director, Nuclear Licensing and Rockville, MD 20852 Operations Support Innsbrook Technical Center Senior Resident Inspector 5000 Dominion Boulevard Millstone Power Station Glen Allen, VA 23060-6711 c/o U.S. Nuclear Regulatory Commission P. O. Box 513 Niantic, CT 06357 Mr. J. W. "Bill" Sheehan Co-Chair NEAC 19 Laurel Crest Drive Waterford, CT 06385 Ms. Nancy Burton 147 Cross Highway Redding Ridge, CT 00870
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS NOS. RR-89-60 AND IR-2-44 MILLSTONE POWER STATION, UNIT NOS. 2 AND 3 DOMINION NUCLEAR CONNECTICUT, INC.
DOCKET NUMBERS 50-336 AND 50-423
1.0 INTRODUCTION
By letter dated May 11, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted to the U.S. Nuclear Regulatory Commission (NRC), a request for approval to use alternatives to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI at the Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3). Specifically, DNC requests the use of Performance Demonstration Initiative (PDI) qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel shell-to-flange weld in accordance with ASME Code,Section XI, Division I,1995 Edition,1996 Addenda, Appendix VIII, Supplements 4 and 6. The relief is requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a(a)(3)(i).
2.0 REGULATORY EVALUATION
The inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by NRC pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall 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 ISI 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(b) twelve months prior to the start of the 120-month interval, subject to the limitations and Enclosure
modifications listed therein. The ISI Code of record for the third 10-year ISI interval for MPS2 and the second 10-year ISI interval for MPS3 is the 1989 Edition of the ASME Code,Section XI. In addition, 10 CFR 50.55a(g)(6)(ii)(C)(2) requires that licensees using the 1989 Edition or earlier editions must implement the 1995 Edition with the 1996 Addenda of Appendix VIII and the supplements to Appendix VIII of Section XI of the ASME Code.
3.0 EVALUATION FOR RELIEF REQUEST NOS. RR-89-60 AND IR-2-44 3.1 Components for Which Relief is Requested ASME Code Class 1, Item B1.30, Reactor Pressure Vessel (RPV) Shell-to-Flange welds identified as:
Millstone 2: Weld FS-1 Millstone 3: Weld 101-121 3.2 Code Requirements The 1989 Edition of ASME Code,Section XI, Appendix I, Subparagraph I-2110, requires that ultrasonic testing (UT) of RPV shell-to-flange welds be conducted in accordance with Article 4 of ASME Code,Section V, supplemented by the requirements of Table I-2000-1. In addition, Regulatory Guide (RG) 1.150, Revision 1, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, serves as guidance for UT examination of RPV welds.
3.3 Licensees Proposed Alternative and Basis for Use During the upcoming 10-year RPV vessel examinations, the licensee proposes to perform ultrasonic examinations of the RPV shell-to-flange welds using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a. The examination will be performed automated as qualified in accordance with ASME Code,Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI Program. Since the subject examinations will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single side access for examination of these welds.
Appendix VIII requirements were developed and adopted to ensure the effectiveness of ultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations. The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to or surpass the requirements of the ASME Code,Section V, Article 4 ultrasonic examination requirements.
The licensee finds that though Appendix VIII is not required for the RPV shell-to-flange weld examination, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing of flaws in the subject welds will be equal to or exceed the requirements of ASME Code,Section V, Article 4 and the guidance in RG 1.150.
3.4 NRC Staff Evaluation The ASME Code requires that ultrasonic examination of shell-to-flange welds in vessels greater than 2 inches in thickness be conducted in accordance with Article 4 of the ASME Code,Section V, as supplemented by requirements in Table I-2000-1. ASME Code,Section V, Article 4 provides a prescriptive process for qualifying UT procedures and performing examinations. The licensee instead proposes to use procedures and personnel qualified in accordance with performance-based criteria listed in the 1995 Edition, 1996 Addenda of the ASME Code,Section XI, Appendix VIII, Supplements 4 and 6 as implemented by the industrys PDI program. These performance-based methods are currently required by 10 CFR 50.55a for examination of all other RPV shell welds (having replaced the Article 4 techniques).
Amplitude-based sizing techniques such as the prescriptive UT procedures that comply with the requirements of Article 4 of ASME Code,Section V, are based on the amplitude of the returned signal and correlating that amplitude with an equivalent machined reflector such as a notch or a side-drilled hole. However, correlation between defect size and amplitude has been poor.
This is not unexpected given the number of variables from the material, equipment and defect itself. The material has potential velocity and microstructure variations, and the equipment has potential amplitude variations due to the type of pulser, frequency band, cabling, and other inherent electrical parameters. Perhaps the biggest variable is the defect itself. Ultrasonic examination is highly sensitive to defect orientation. Also, transparency, roughness, curvature, and location play a role in the ability to detect and size defects. In addition, conventional amplitude-based ultrasonics is particularly unreliable for vertical defects.
When prescriptive UT procedures that comply with the requirements of Article 4 of ASME Code,Section V, were used in round robin tests containing real flaws in RPV mockups, and the results statistically analyzed according to the screening criteria of ASME Code,Section XI, Appendix VIII, the procedures proved to be less effective than examinations that utilize Appendix VIII, Supplements 4 and 6, qualified procedures. Performance-based UT is generally applied with higher sensitivity, which increases the probability of detecting a flaw when compared to prescriptive Section V, Article 4 requirements. Also, flaw sizing is more accurately determined with the time-based tip diffraction criteria used by performance-based ultrasonics than with the less accurate amplitude criteria for prescriptive Section V, Article 4 requirement.
Procedures, equipment, and personnel qualified through the PDI program have demonstrated their skill level to detect flaws common to nuclear power plants and have shown high probability of detection levels. This has resulted in an increased reliability of inspections for weld configurations subject to the requirements of Appendix VIII.
The licensee states that the examination, using automated equipment, will be performed by procedures qualified in accordance with ASME Code,Section XI, 1995 Edition with 1996 Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI program. Due to the weld configuration the examination will be performed from a single side.
Therefore, all procedures, personnel, and equipment will be qualified for single-sided access for examination of the subject welds.
3.5 Summary The NRC staff finds that the use of UT procedures and personnel qualified to the 1995 Edition with 1996 Addenda of Section XI of the ASME Code, Appendix VIII, Supplement 4 and 6, as modified by 10 CFR 50.55a by demonstration through the PDI program for the RPV shell-to-flange weld, provides equivalent or better examination results than those obtained from ASME Code,Section V requirements and RG 1.150 recommendations. Therefore, based on the above analysis, the staff considers that an acceptable level of quality and safety will be maintained when using the licensees proposed alternative examination.
4.0 CONCLUSION
The NRC staff concludes that the proposed alternative with PDI qualified procedures and personnel applied from the RPV shell surface along with the improved capabilities as discussed above will provide equivalent or better examination results than those realized from the ASME Code,Section V, requirements and RG 1.150 recommendations and, therefore, will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC authorizes the proposed alternative for the remainder of the third 10-year ISI interval at MPS2 and the second 10-year ISI interval at MPS3.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this, or previous, relief requests remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: A. Keim Date: May 1, 2007
