|
---|
Category:Letter
MONTHYEARIR 05000282/20230042024-02-0101 February 2024 Integrated Inspection Report 05000282/2023004 and 05000306/2023004 ML23356A1232024-01-29029 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) ML24024A0722024-01-24024 January 2024 Independent Spent Fuel Storage Installation, Onticello, Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML24017A0182024-01-19019 January 2024 Confirmation of Initial License Examination ML23356A0032024-01-17017 January 2024 Issuance of Amendments Revise Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report IR 07200010/20234012023-12-20020 December 2023 Independent Spent Fuel Storage Installation Security Inspection Report 07200010/2023401 ML23349A0572023-12-15015 December 2023 and Independent Spent Fuel Storage Installation, Revision to Correspondence Service List for Northern States Power - Minnesota ML23215A1672023-12-15015 December 2023 Acceptance of Requested Licensing Action Amendment Request to Revise Surveillance Requirement 3.8.1.2 Note 3 IR 05000282/20234012023-12-13013 December 2023 Security Baseline Inspection Report 05000282/2023401 and 05000306/2023401 L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 ML23304A1632023-11-15015 November 2023 Supplemental Information Needed for Acceptance of Requested Licensing Action Amendment Request to Revise SR 3.8.1.2 Note 3 ML23319A3182023-11-15015 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000282/20230032023-11-0808 November 2023 Integrated Inspection Report 05000282/2023003 and 05000306/2023003 ML23311A3572023-11-0707 November 2023 Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant (PINGP) Unit 2. Cycle 33. Revision 0 ML23285A3062023-10-12012 October 2023 Implementation of the Fleet Standard Emergency Plan for the Monticello Nuclear Generating Plant and the Prairie Island Nuclear Generating Plant ML23270B9022023-09-29029 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 ML23265A2532023-09-26026 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23262B0372023-09-19019 September 2023 Response to NRC Request for Additional Information Regarding the 2023 Monticello and Prairie Island Plant Decommissioning Funding Status Reports ML23256A1682023-09-13013 September 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Voluntary Security Clearance Program 2023 Insider Threat Program Self-Inspection L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy IR 05000282/20230052023-08-30030 August 2023 Updated Inspected Plan for Prairie Island Nuclear Generating Plant Report 05000282/2023005 and 05000306/2023005 IR 05000282/20230102023-08-17017 August 2023 NRC Inspection Report 05000282/2023010 and 05000306/2023010 ML23222A0122023-08-10010 August 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Changes in Foreign Ownership, Control or Influence IR 05000282/20230022023-08-0303 August 2023 Integrated Inspection Report 05000282/2023002 and 05000306/2023002 ML23214A2032023-08-0202 August 2023 Request for Information for an NRC Quadrennial Comprehensive Engineering Team Inspection: Inspection Report 05000282/2024010; 05000306/2024010 ML23206A2342023-07-25025 July 2023 Independent Spent Fuel Storage Installation, and Monticello Nuclear Generating Plant, Changes in Foreign Ownership, Control or Influence ML23202A0032023-07-21021 July 2023 Independent Spent Fuel and Independent Spent Fuel Storage Installation, Monticello Nuclear Generating Plant, Submittal of Quality Assurance Topical Report (NSPM-1) ML23199A0922023-07-18018 July 2023 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000306/2023004 ML23195A1732023-07-14014 July 2023 Revision of Standard Practice Procedures Plan L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT ML23181A0192023-06-30030 June 2023 Independent Spent Fuel Storage Installation, Revision to Correspondence Service List for Northern States Power - Minnesota ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report IR 05000282/20234202023-06-0101 June 2023 Security Baseline Inspection Report 05000282/2023420 and 05000306/2023420 ML23150A1722023-05-30030 May 2023 Preparation and Scheduling of Operator Licensing Examinations 2024-02-01
[Table view] Category:Report
MONTHYEARML23270B9022023-09-29029 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23265A2532023-09-26026 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23075A3512022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 6 of 7) ML23075A3482022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 3 of 7) ML23075A3472022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 2 of 7) L-PI-23-002, WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 7 of 7)2022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 7 of 7) ML23075A3462022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 1 of 7) ML23075A3502022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 5 of 7) ML23075A3492022-12-31031 December 2022 WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, (Part 4 of 7) ML22003A1832022-01-0303 January 2022 Refueling Outage Unit 2 R32 Owner'S Activity Report for Class 1, 2, 3 and Mc Inservice Inspections ML21285A2812021-10-12012 October 2021 Revised Pressure and Temperature Limits Report ML20272A2932020-09-28028 September 2020 (PINGP) Unit 1 and 2 Revised Pressure and Temperature Limits Report ML20265A0892020-09-15015 September 2020 Draft License Conversation Record L-PI-19-040, License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency2019-10-0707 October 2019 License Amendment Request: Revise Technical Specification 5.5.14 to Permanently Extend Containment Leakage Rate Test Frequency ML17279A1242017-09-30030 September 2017 Enclosure 5 to L-PI-17-041, Westinghouse WCAP-17400-NP, Supplemental 1, Revision 2, Spent Fuel Pool Criticality Safety Analysis Supplemental Analysis Including the Storage of Ifba Bearing Fuel L-PI-16-058, Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation2016-07-22022 July 2016 Special Report: Timely Restoration of Operability of Explosive Gas Monitoring Instrumentation L-PI-16-054, Pressure and Temperature Limits Report, Revision 52016-06-22022 June 2016 Pressure and Temperature Limits Report, Revision 5 L-PI-16-051, 10 CFR 50.46 Emergency Core Cooling System Annual Report2016-06-22022 June 2016 10 CFR 50.46 Emergency Core Cooling System Annual Report L-PI-15-034, Pressure and Temperature Limits Report (PTLR) Revision 42015-05-14014 May 2015 Pressure and Temperature Limits Report (PTLR) Revision 4 ML15037A4582015-03-0606 March 2015 Staff Assessment of the Aging Management Program for Reactor Vessel Internal Components L-PI-14-131, Fifth Ten-Year Interval Snubbers Testing Program2014-12-18018 December 2014 Fifth Ten-Year Interval Snubbers Testing Program ML17297A3232014-11-14014 November 2014 Enclosure 2 (Redacted): Seismic Walkdown Report, in Response to the 50.54(f) Information Request Re Fukushima Near-Term Task Force Recommendation 2.3: Seismic Updated Transmittal for Prairie Island Unit 1 ML16005A1102014-09-25025 September 2014 Redacted 2014 Decommissioning Cost Analysis for the Prairie Island Nuclear Generating Plant ML14148A4772014-06-17017 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident ML14120A1622014-05-0909 May 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident L-PI-14-045, Enclosure to L-PI-14-045, Transition Report, Revision 12014-04-30030 April 2014 Enclosure to L-PI-14-045, Transition Report, Revision 1 L-PI-14-028, PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2014-03-27027 March 2014 PINGP Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML14030A5402014-02-27027 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14041A2042014-02-26026 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Prairie Island Nuclear Generating Plant, Units 1 and 2, TAC Nos.: MF0834 and MF0835 L-PI-13-080, First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2013-08-26026 August 2013 First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events L-PI-12-108, Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Information, Per 10CFR50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident L-PI-12-103, Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident2012-11-26026 November 2012 Final Response to NRC Request for Lnformation Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Lnsiqhts from the Fukushima Dai-ichi Accident ML12278A4052012-09-28028 September 2012 Prairie Island, Units 1 and 2, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors ML13133A0632012-06-27027 June 2012 H4, Rev. 27, Offsite Dose Calculation Manual (Odcm). ML12159A2562012-06-11011 June 2012 Review of 60-Day Response to Request for Information Regarding Recommendation 9.3 of the Near-Term Task Force Related to the Fukushima Daiichi Nuclear Power Plant Accident L-PI-10-076, Calculation No. GEN-PI-083, Revision 1, Locked Rotor Accident (LRA) Analysis Using AST, Attachment 52010-07-23023 July 2010 Calculation No. GEN-PI-083, Revision 1, Locked Rotor Accident (LRA) Analysis Using AST, Attachment 5 ML1021002592010-06-30030 June 2010 Seismic Fragilities for Unit #1 and Unit #2 Turbine Building Piping and Equipment ML1016901712010-06-11011 June 2010 Enclosure 6, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16275, Effects of Pipe Whip Interactions for Various Pipe Combinations for Internal Flooding Sdp. ML1016901702010-06-11011 June 2010 Enclosure 5, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16270, Screening of Pipe Whip Interactions for Sdp. ML1016901682010-06-11011 June 2010 Enclosure 3, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16090, Turbine Building Flooding SDP: Cl Turbine Building Pipe Break Analysis. ML1016901692010-06-10010 June 2010 Enclosure 4, Prairie Island Nuclear Generating Plant, Units 1 and 2, Supporting Engineering Evaluations, EC 16154, Turbine Building Flooding SDP: Cl Turbine Building Seismic Pipe Break Analysis. L-PI-10-005, Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems2010-02-18018 February 2010 Ninety-Day 1R26 Post-Outage Report Pursuant to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal & Containment Spray Systems ML1019703862009-12-31031 December 2009 Ground Water Investigation: an Improved Flow Net to Evaluate Pathways for a Potential Ground Water Release ML1002001312009-12-21021 December 2009 Report No. 0900634.401, Revision 2, Updated Leak-Before-Break Evaluation for Several RCS Piping at Prairie Island Nuclear Generating Plant, Units 1 & 2. ML1002001322009-12-18018 December 2009 Report 0900634.402, Revision 2, Updated Leak-Before-Break Report for Prairie Island Nuclear Generating Plant Unit 2 Pressurizer Surge Line Nozzle. L-PI-09-115, Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology2009-10-27027 October 2009 Radtrad and ARCON96 Input and Output Files in Support of License Amendment Request (LAR) to Adopt the Alternative Source Term Methodology ML1008406402009-09-11011 September 2009 Advisory Brief of Prairie Island Nuclear Generating Plant Study Group to State of Minnesota, Office of Administrative Hearings for the Public Utilities Commission, Sept. 11, 2009. Submitted with Comments on Draft Generic Environmental Impac L-PI-09-021, 2008 Unit 2 180-Day Steam Generator Tube Inspection Report2009-04-27027 April 2009 2008 Unit 2 180-Day Steam Generator Tube Inspection Report ML1020302362008-12-31031 December 2008 State of Wisconsin Prairie Island Environmental Radioactivity Survey ML0834701962008-11-21021 November 2008 PINGP - License Renewal; Radon Health Risks 2023-09-29
[Table view]Some use of "" in your query was not closed by a matching "". Category:Miscellaneous
Some use of "" in your query was not closed by a matching "".
[Table view]Some use of "" in your query was not closed by a matching "". |
Text
September 29, 2023 Mr. Thomas A. Conboy Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 - REVIEW OF REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM CAPSULE N TECHNICAL REPORT (EPID L-2023-LRO-0010)
Dear Mr. Conboy:
By letter dated March 16, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23075A345), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (NSPM, the licensee) submitted report WCAP-18795-NP, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program. The report was provided in accordance with Title 10 of the Code of Federal Regulations (10 CFR), part 50, appendix H, section IV. Testing was performed in accordance with American Society for Testing and Materials (ASTM) Standard E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, as specified in 10 CFR, part 50, appendix H, paragraph III.B.1.
The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of NSPMs submittal as documented in the enclosed staff evaluation. The NRC staff concludes that NSPM has provided the information required by the regulations and that no additional follow-up is required at this time. This completes the NRC staffs efforts for EPID L-2023-LRO-0010.
T. Conboy If you have any questions, please contact me at 301-415-0680 or via e-mail at Brent.Ballard@nrc.gov.
Sincerely,
/RA/
Brent T. Ballard, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-306
Enclosure:
Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report cc: Listserv
OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM CAPSULE N TECHNICAL REPORT NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT 2 DOCKET NO. 50-306
1.0 INTRODUCTION
By letter dated March 16, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23075A344), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (NSPM, the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) an evaluation of the testing results of reactor vessel radiation surveillance program Capsule N for the Prairie Island Nuclear Generating Plant (Prairie Island),
Unit 2, in accordance with the Title 10 of the Code of Federal Regulations (10 CFR), part 50, appendix H, Reactor Vessel Material Surveillance Program Requirements. The evaluation report is titled WCAP-18795-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, dated December 2022.
2.0 REGULATORY EVALUATION
The regulations in 10 CFR, part 50, appendix H, requires licensees to implement a material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light-water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment.
Paragraph IV.A of appendix H to 10 CFR, part 50, specifies that a summary technical report for each capsule withdrawal and the associated test results must be submitted within 18 months of the date of capsule withdrawal, unless an extension is granted by the Director, Office of Nuclear Reactor Regulation.
Paragraph IV.B of appendix H to 10 CFR, part 50, requires that capsule evaluation reports include all data specified by American Society for Testing and Materials (ASTM) Standard Practice E185-82 and the results of all fracture toughness tests conducted on the surveillance capsule materials in both the unirradiated and irradiated condition.
Paragraph IV.C of appendix H to 10 CFR, part 50, requires that if a change in the technical specifications (TSs) is required, either in the pressure-temperature (P/T) limits or in the operating procedures required to meet the limits, the expected date for submittal of the revised TSs must be provided with the report.
Enclosure
The NRC Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988 (ML003740284), provides guidance on general procedures acceptable to the NRC staff for calculating effects of neutron radiation embrittlement of low-alloy steels used for light-water-cooled reactor vessels.
3.0 NRC STAFF EVALUATION 3.1 Surveillance Capsule Program Irradiation surveillance of the reactor vessel is necessary to assure that the vessel material will maintain its fracture toughness throughout the service life of the plant. The surveillance capsule contains both dosimeters as well as archival material samples to be irradiated to levels comparable to those expected to be accrued by the reactor vessel at the end of its licensed period. Under the program, fracture toughness test data is obtained from the material specimens exposed in the surveillance capsules which are withdrawn periodically from the reactor vessel.
The Prairie Island, Unit 2, reactor pressure vessel (RPV) material surveillance program was developed based on ASTM E185-70 Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. The program contains a total of six capsules. Capsule N is the fifth capsule that has been removed. The licensee removed Capsule N from the reactor vessel at 40.64 effective full-power years (EFPY).
3.2 Neutron Fluence Evaluation A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-VI). The licensee reported that Capsule N was removed at 40.64 EFPY and received a fluence of 8.41 x 1019 n/cm2 (newtons per square centimeter) (E >1.0 MeV).
In WCAP-18795-NP, Revision 0, the licensee stated that the transport calculations supporting the analysis of the fluence of Capsule N were carried out using the RAPTOR-M3G computer code. While RAPTOR-M3G has been reviewed and approved by the NRC staff for referencing in licensing applications via WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET (ML18204A010), the licensee has not been approved by the NRC to incorporate this methodology into the licensing basis for Prairie Island, Unit 2. From the NRC staffs review of appendix A of the licensees submittal, which compares the measured capsule dosimetry results with those of previously withdrawn capsules, the NRC finds that no safety significant issue is presented or implied by the results of the capsule surveillance report. This statement should not be construed as NRC approval for use of RAPTOR-M3G as a method for evaluation for Prairie Island, Unit 2. The use of RAPTOR-M3G would require appropriate incorporation of the methodology into the Prairie Island, Unit 2, licensing basis.
The licensee projected the peak clad/base metal interface vessel fluence at 54 EFPY (end-of-license extension) of plant operation to be 5.66 x 1019 n/cm2 (E > 1.0 MeV). Using the RG 1.99, Revision 2, attenuation formula and a vessel thickness of 6.692 inches, the licensee calculated vessel peak quarter-thickness (1/4T) and three-quarter thickness (3/4T) fluence of 3.79 x 1019 n/cm2 and 1.70 x 1019 n/cm2, respectively.
3.3 Material Test Results The licensee performed mechanical tests of the Charpy V-notch and tensile specimens in Capsule N in accordance with 10 CFR, part 50, appendix H, and ASTM Specification E185-82.
3.3.1 Transition Temperature Shift The reactor vessel lower shell forging D (Heat # 22642) Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation),
resulted in an irradiated 30 ft-lb (foot-pound) transition temperature of 148.2 °F (degree Fahrenheit). This results in a 30 ft-lb transition temperature increase of 176.5 °F.
The reactor vessel lower shell forging D (Heat # 22642) Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (axial orientation),
resulted in an irradiated 30 ft-lb transition temperature of 152.9 °F. This results in a 30 ft-lb transition temperature increase of 154.1 °F.
The weld material (Heat # 2721, Flux Type UM89, Lot # 1263) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 59.5 °F. This results in a 30 ft-lb transition temperature increase of 135.6 °F.
The reactor vessel heat-affected zone (HAZMAT) material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 10.9 °F. This results in a 30 ft-lb transition temperature increase of 145.5 °F.
The licensees test results show that the measured shifts in the 30 ft-lb transition temperature of all the surveillance materials are higher than the RG 1.99, Revision 2, position 1.1, predictions.
The test results show that the RG 1.99, Revision 2, prediction is not accurate and not conservative when the RPV material is irradiated to a high neutron fluence. The NRC staff will consider the implications of the non-conservativism in future licensing actions, such as when the licensee evaluates reactor pressure vessel embrittlement for license renewal.
3.3.2 Upper Shelf Energy The average upper-shelf energy of lower shell forging D (Heat # 22642) (tangential orientation) resulted in an average energy decrease of 22.7 ft-lb. This decrease results in an irradiated average upper-shelf energy of 125 ft-lb for the tangentially oriented specimens.
The average upper-shelf energy of lower shell forging D (Heat # 22642) (axial orientation) resulted in an average energy decrease of 17.2 ft-lb. This decrease results in an irradiated average upper-shelf energy of 89 ft-lb for the axially oriented specimens.
The average upper-shelf energy of the surveillance program weld material (Heat # 2721)
Charpy specimens resulted in an average energy decrease of 8.3 ft-lb. This decrease results in an irradiated average upper-shelf energy of 95 ft-lb.
The average upper-shelf energy of the HAZMAT material Charpy specimens resulted in an average energy decrease of 32 ft-lb. This decrease results in an irradiated average upper-shelf energy of 82 ft-lb.
The measured percent decreases in upper-shelf energy of all the surveillance materials, i.e., the Lower Shell Forging D (Heat # 22642) and weld material (Heat # 2721) are less than the RG 1.99, Revision 2, position 1.2 predictions. These results show that RG 1.99 method is conservative in the predicting the upper-shelf energy of the reactor vessel material.
3.3.3 Credibility Evaluation The NRC staff notes that the credibility assessment performed in appendix D of WCAP-18795-NP, Revision 0, found the surveillance weld data to be credible and the forging data to be non-credible in accordance with the RG 1.99, Revision 2. The licensee stated in its evaluation that although the lower shell forging D did not meet Criterion 3, both materials may still be used in determining the upper-shelf energy decrease in accordance with RG 1.99, Revision 2, position 2.2. The NRC staff reviewed the credibility evaluation and had no objections. However, surveillance data determined to be non-credible may still need to be factored into licensing calculations, such as P/T limits.
4.0 CONCLUSION
The licensee is expected to incorporate the updated surveillance data into the next revision of the Prairie Island, Unit 2, P/T limits report, as required by 10 CFR, part 50, appendix G. The licensee specified an expected submittal date in their submission and has submitted a license amendment request (LAR) to revise Prairie Island Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to incorporate use of newer analytical methods to determine the RCS P/T limits and cold overpressure mitigation system setpoints (ML22343A257). The LAR is currently under review by NRC staff.
The NRC staff finds that the licensees report of reactor vessel surveillance capsule N from Unit 2 satisfies the requirements of 10 CFR, part 50, appendix H, and that the licensee performed tests and calculations based on 10 CFR 50, appendix H and RG 1.99, Revision 2, appropriately. The staff also concludes that the licensees submittal meets the reporting requirements in section IV of 10 CFR, part 50, appendix H.
Principal Contributors: MBenson, NRR JTsao, NRR SBhatt, NRR Date: September 29, 2023
ML23270B902 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NAME BBallard SRohrer PSahd DATE 9/26/2023 9/27/23 7/18/2023 OFFICE NRR/DNRL/NVIB/BC NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME ABuford (DWidrevitz for) JWhited BBallard DATE 9/14/2023 9/28/2023 9/29/2023