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MONTHYEARML1027800862010-10-27027 October 2010 Issuance of Relief Request RR-89-68 Regarding Use of American Society of Mechanical Engineering Code Case N-722-1 Project stage: Approval 2010-10-27
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Category:Letter
MONTHYEARIR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 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 ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations 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 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) 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) ML24142A0952024-05-20020 May 2024 End of Cycle 22 Steam Generator Tube Inspection Report 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 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 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) IR 05000336/20244402024-04-25025 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 ML24086A4762024-03-22022 March 2024 Application for Technical Specification Change to Extend the Inspection Interval for Reactor Coolant Pump Flywheels Using the Consolidated Line-Item Improvement Process ML24086A4802024-03-22022 March 2024 Alternative Request IR-4-14, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examination for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles ML24051A1922024-03-0808 March 2024 – Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) IR 05000336/20230062024-02-28028 February 2024 Annual Assessment Letter for Millstone Power Station, Units 2 and 3, (Reports 05000336/2023006 and 05000423/2023006) ML24053A2632024-02-21021 February 2024 Unit 3, and Independent Spent Fuel Storage Installation, Notification Pursuant to 10 CFR 72.212(b)(1) Prior to First Storage of Spent Fuel Under a General License ML24057A0612024-02-19019 February 2024 and Virgil C. Summer Power Nuclear Stations - Nuclear Property Insurance Coverage ML24043A0062024-02-15015 February 2024 Request for Withholding Information from Public Disclosure for License Amendment Request to Support Implementation of Framatome Gaia Fuel IR 05000336/20230042024-02-14014 February 2024 Integrated Inspection Report 05000336/2023004 and 05000423/2023004 ML24030A7522024-01-30030 January 2024 Technical Specification Bases Pages IR 05000336/20234022024-01-30030 January 2024 Security Baseline Inspection Report 05000336/2023402 and 05000423/2023402 (Cover Letter Only) 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 IR 05000336/20234402024-01-11011 January 2024 Special Inspection Report 05000336/2023440 and 05000423/2023440 (Cover Letter Only) ML24004A1052024-01-0404 January 2024 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000336/2024010 & 05000423/2024010 ML23361A0942023-12-21021 December 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies and Core Operating Limits Report . 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) ML23361A0312023-12-20020 December 2023 Intent to Pursue Subsequent License Renewal ML23352A0202023-12-18018 December 2023 Senior Reactor and Reactor Operator Initial License Examinations 2024-08-06
[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 ML19338G3722019-12-18018 December 2019 Alternative Requests RR-05-01 and RR-05-02 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 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
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 27, 2010 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 SUB~IECT: MILLSTONE POWER STATION, UNIT NO.2-ISSUANCE OF RELIEF REQUEST RR-89-68 REGARDING USE OF AMERICAN SOCIETY OF MECHANICAL ENGINEERING CODE CASE N-722-1 (TAC NO. ME2309)
Dear Mr. Heacock:
By letter dated October 1, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092740662), Dominion Nuclear Connecticut, Inc. (DNC or the licensee), submitted relief request RR-89-68 requesting to use the alternative requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)
Case N-722-1 when implementing Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(6)(ii)(E) for Millstone Power Station, Unit NO.2 (MPS2). The proposed alternative would eliminate the need for the visual examinations of a series of welds in the reactor pressure vessel (RPV) flange seal leak-off line at MPS2, for the third 10-year inservice inspection (lSI) interval.
The Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the proposed use of ASME Code Case N-722-1 for welds in the RPV flange seal leak-off line at MPS2 provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC authorizes the use of RR-89-68 for the third 1O-year interval at MPS2. The third 10-year lSI concluded on March 31,2010.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the authorized Nuclear Inservice Inspector.
D. Heacock -2 If you have any questions, please contact the Project Manager, Carleen Sanders, at 301-415-1603.
sin.!.c2(/ r(/~
f IC 4arold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosure:
As stated cc wI encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST RR-89-68 MILLSTONE POWER STATION, UNIT NO.2 DOMINION NUCLEAR CONNECTICUT, INC.
DOCKET NUMBER 50-336
1.0 INTRODUCTION
By letter dated, October 1, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092740662), Dominion Nuclear Connecticut, Inc. (DNC or the licensee), submitted relief request RR-89-68 requesting to use the alternative requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)
Case N-722-1 when implementing Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(6)(ii)(E) for Millstone Power Station, Unit NO.2 (MPS2). The proposed alternative would eliminate the need for the visual examinations of a series of welds in the reactor pressure vessel (RPV) flange seal leak-off line at MPS2, for the third 10-year in-service inspection (lSI) interval.
2.0 REGULATORY EVALUATION
The lSI of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states 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 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 1O-year interval and subsequent intervals comply with the requirements in the latest edition Enclosure
-2 and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ASME Code of Record for the MPS2 third 1O-year lSI interval is the 1989 Edition with no Addenda.
10 CFR 50.55a(g)(6)(ii) states that the Commission may require the licensee to follow an augmented lSI program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. 10 CFR 50.55a(g)(6)(ii)(E) requires augmented lSI of reactor vessel bottom-mounted instrumentation penetration nozzles of pressurized-water reactors in accordance with ASME Code Case N-722, subject to the conditions specified in paragraphs (2) through (4) of 10 CFR 50.55a(g)(6)(ii)(E).
DNC requests relief from certain requirements of 10 CFR 50.55a(g)(6)(ii)(E) for the third 10-year lSI interval at MPS2 under 10 CFR 50.55a(a)(3).
3.0 TECHNICAL EVALUATION
3.1 Affected Systems and Components The items for which relief is requested are three Alloy 600/82/182 welds classified as ASME Code Class 1, B15.100 instrument connections in the RPV flange seal leak-off line monitoring tube. These welds are incased within the RPV flange. The RPV flange seal leak-off line is a combination of a %" SB-166 Alloy 600 tube and a %" SB-167 Alloy 600 pipe. The three connections are a partial penetration weld on the flange inside surface, a butt weld joining the tube-to-pipe sections, and a socket weld joining the pipe to a stainless steel elbow at the end of the pipe section.
3.2 Proposed Alternative DNC proposes to use ASME Code-approved Code Case N-722-1 in lieu of Code Case N-722 for the requirements associated with the Class 1 welds made of Alloy 600/82/182 materials in the RPV flange seal leak-off line at MPS2.
3.3 Licensee's Basis for Request The RPV flange seal leak-off line is used to support the identification of inner O-ring seal leakage. This system is essentially a drain system. During normal operation, this leak-off line does not experience flow and is not pressurized. Therefore, these welds would not normally be exposed to an environment that could cause primary water stress-corrosion cracking (PWSCC);
the identification of which is the primary focus of ASME Code Case N-722. Therefore, it is reasonable that the RPV flange seal leak-off line does not belong in the same scope of piping and components that are included in ASME Code Case N-722.
In accordance with 10 CFR 50.55a(g)(6)(ii)(E), MPS2 must follow ASME Code Case N-722 for the Alloy 600/82/182 ASME Code Class 1 B15.100 instrument connection in the RPV flange seal leak-off line. Code Case N-722 requires the welds in the RPV seal flange leak-off line be visually inspected once per 10-year interval. The welds in this system are not accessible for direct bare metal visual inspection. Instead, DNC proposes to use ASME Code Case N-722-1.
ASME Code Case N-722-1 is a revision to ASME Code Case N-722 that was approved for use
-3 by the ASME Code on January 26,2009. ASME Code Case N-722-1 does not require visual examinations of the welds in the RPV flange seal leak-off line. This proposed alternative has an acceptable level of quality and safety because the welds in the RPV flange seal leak-off line are not normally subject to continuous contact with reactor coolant during normal operation, and are not subject to PWSCC concerns. Therefore, this proposed alternative provides an acceptable level of quality and safety equal to what can be gained by performing examinations of these welds to the current Code Case N-722 requirements.
3.4 NRC Staff's Evaluation The NRC staff's review of this request was based on 10 CFR 50.55a(a)(3)(i) which states that:
The proposed alternatives would provide an acceptable level of quality and safety On September 10, 2008, the Commission added new requirements in paragraph 50.55a(g)(6)(ii)(E) that pertain to reactor coolant pressure boundary visual inspections. In accordance with 50.55a(g)(6)(ii)(E), the visual inspection requirements for Item Number B15.100, Reactor Vessel Instrument Connections, must be performed once per interval.
The intention of ASME Code Case N-722 was to provide a bare metal visual inspection frequency for Alloy 600/82/182 components in U.S. pressurized-water reactors, with the exception of the upper head penetration nozzles, as these inspections are addressed in another ASME Code Case (Code Case N-729-1). The basis for the inspection is to identify PWSCC. In order for PWSCC to initiate in a component, it requires a susceptible material, a tensile driving stress and a suitable environment. Alloy 600/82/182 material in contact with primary coolant provides a suitable environment for PWSCC. Although the weld material is considered susceptible and a tensile driving stress could exist, without a suitable environment the NRC staff considers PWSCC unlikely. The generic scope of terminology used in ASME Code Case N-722, "instrument connections" to the RPV that use Alloy 600/82/182 materials, has included the instrument connection welds of the RPV flange seal leak-off line which are not normally exposed to primary coolant and therefore are not normally susceptible to PWSCC. Therefore, the NRC staff finds that the RPV flange seal leak-off line instrument connections at MPS2 do not need to be a part of this additional augmented inspection requirement.
ASME-approved Code Case N-722-1 does not require visual examination of the RPV flange seal leak-off line. Code Case N-722-1 is currently under rule making, but is not currently approved for use by the NRC staff. However, based on the information provided above, the NRC staff finds that the licensee's proposed alternative to use the inspection requirements of ASME Code Case N-722-1 in lieu of N-722 for the RPV flange seal leak-off line when implementing the requirements of 10 CFR 50.55a(g)(6)(ii)(E) provides an acceptable level of quality and safety. Therefore, the NRC staff finds sufficient basis to authorize relief under 10 CFR 50.55a(a)(3)(i).
4.0 CONCLUSION
Based on the discussions above, the NRC staff concludes that the use of Code Case N-722-1 for the RPV flange seal leak-off line welds meets the intention of 50.55a(g)(6)(ii)(E). The NRC staff also concludes, based on the above discussion, that RR 89-68 provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff
-4 authorizes the use of relief request RR 89-68 for the third 10-year lSI interval at MPS2. The third 10-year lSI concluded on March 31, 2010.
All other ASME Code,Section XI, requirements for which relief has not been specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Collins Date: October 27,2010
D. Heacock -2 If you have any questions, please contact the Project Manager, Carleen Sanders, at 301-415-1603.
Sincerely,
/raJ Harold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosure:
As stated cc wi end: Distribution via Listserv Distribution:
PUBLIC Branch Reading RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl1-2 Resource RidsNrrDciCpnb Resource RidsNrrLAABaxter Resource RidsNrrMillstone Resource RidsOgcMailCenter RidsRgn1 MailCenter Resource RidsSecyMailCenter Resource SWilliams, EDO R-1 JCollins, NRR ADAMS Accession N0.: ML102780086 *Bsy Memo Dated OFFICE LPL 1-2/PM LPL 1-2/LA DCI/CPNB LPL 1-2/BC NAME CSanders ABaxter TLupold* HChernoff (w/comment)
DATE 10/18/10 10/18/10 09/02/2010 10/27/10 10/27/10 OFFICAL RECORD COpy