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MONTHYEARML1002001312009-12-21021 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. Project stage: Request ML1005701342010-02-25025 February 2010 Acceptance Review of LAR to Apply Leak-Before-Break Methodology (TAC Nos. ME2976 and ME2977 Project stage: Acceptance Review ML1011802112010-05-0606 May 2010 Request for Withholding Information from Public Disclosure for Prairie Island Nuclear Generating Plant, Units 1 and 2 Project stage: Withholding Request Acceptance ML1011708332010-05-0606 May 2010 Request for Withholding Information from Public Disclosure for Prairie Island Nuclear Generating Plant, Units 1 and 2 Project stage: Withholding Request Acceptance ML1011708142010-05-0606 May 2010 Request for Withholding Information from Public Disclosure for Prairie Island Nuclear Generation Plant, Units 1 and 2 Project stage: Withholding Request Acceptance ML1015506682010-06-10010 June 2010 Request for Additional Information Related to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-before-Break Project stage: RAI L-PI-10-077, Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology Additional ...2010-07-23023 July 2010 Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology Additional ... Project stage: Supplement L-PI-10-085, Clarification of Responses to Requests for Additional Information Regarding a License Amendment Request for Certain Applications of Leak-Before-Break Methodology2010-08-20020 August 2010 Clarification of Responses to Requests for Additional Information Regarding a License Amendment Request for Certain Applications of Leak-Before-Break Methodology Project stage: Response to RAI L-PI-10-094, Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI2010-10-0808 October 2010 Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI Project stage: Supplement ML1032803982010-12-14014 December 2010 RAI, Related to Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of LBB Methodology Project stage: RAI L-PI-11-006, Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI2011-01-14014 January 2011 Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI Project stage: Supplement L-PI-11-019, Supplement to License Amendment Request to Exclude Dynamic Effects Associated with Certain Postulated Pipe Ruptures from Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to Requests for Clarification2011-02-23023 February 2011 Supplement to License Amendment Request to Exclude Dynamic Effects Associated with Certain Postulated Pipe Ruptures from Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to Requests for Clarification Project stage: Supplement L-PI-11-038, Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology2011-04-0606 April 2011 Supplement to License Amendment Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology Project stage: Supplement ML1120106962011-07-22022 July 2011 RAI, Related to Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology Project stage: RAI L-PI-11-070, Supplement to License Amendment Request to Exclude Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI2011-08-0909 August 2011 Supplement to License Amendment Request to Exclude Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology - Response to RAI Project stage: Supplement ML1122008562011-10-27027 October 2011 Operating Plant, Units 1 and 2 - Issuance of Amendments Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis Based Upon Application of Leak-Before-Break Methodology Project stage: Approval 2010-07-23
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Category:Letter
MONTHYEARIR 05000282/20240052024-08-28028 August 2024 Updated Inspection Plan and Assessment Follow-Up Letter for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Report 05000282/2024005 and 05000306/2024005) IR 05000282/20245012024-08-0505 August 2024 Emergency Preparedness Inspection Report 05000282/2024501 and 05000306/2024501 ML24213A1592024-07-31031 July 2024 Operator Licensing Examination Approval - Prairie Island Nuclear Generating Plant IR 05000282/20240022024-07-30030 July 2024 Integrated Inspection Report 05000282/2024002 and 05000306/2024002 ML24208A1502024-07-26026 July 2024 Independent Spent Fuel Storage Installation - Submittal of Quality Assurance Topical Report (NSPM-1) ML24197A2012024-07-15015 July 2024 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000282/2024004 IR 05000282/20240102024-06-28028 June 2024 Comprehensive Engineering Team Inspection Report 05000282/2024010 and 05000306/2024010 L-PI-24-036, – Preparation and Scheduling of Operator Licensing Examinations2024-06-28028 June 2024 – Preparation and Scheduling of Operator Licensing Examinations ML24158A5912024-06-0606 June 2024 CFR 50.46 LOCA Annual Report L-PI-24-031, Independent Spent Fuel Storage Installation, Supplement to License Amendment Request to Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2)2024-06-0505 June 2024 Independent Spent Fuel Storage Installation, Supplement to License Amendment Request to Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2) L-PI-24-014, License Amendment Request to Revise the Technical Specification Definition of Reactor Trip System (RTS) Response Time and Apply Response Time Testing to RTS Trip Functions with Time Delay Assumption2024-06-0303 June 2024 License Amendment Request to Revise the Technical Specification Definition of Reactor Trip System (RTS) Response Time and Apply Response Time Testing to RTS Trip Functions with Time Delay Assumption ML24155A1922024-05-31031 May 2024 Refueling Outage Unit 2 R33 Owners Activity Report for Class 1, 2, 3 and Mc Inservice Inspections ML24149A3712024-05-29029 May 2024 (Ping) - Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection L-PI-24-030, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.12024-05-22022 May 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.1 ML24141A1292024-05-22022 May 2024 Northern States Power Company - Use of Encryption Software for Electronic Transmission of Safeguards Information ML24141A0452024-05-20020 May 2024 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML24128A2572024-05-16016 May 2024 ISFSI A13 Acceptance Letter IR 05000282/20240012024-05-15015 May 2024 Integrated Inspection Report 05000282/2024001 and 05000306/2024001 ML24130A2392024-05-0909 May 2024 2023 Annual Radioactive Effluent Report ML24130A2362024-05-0909 May 2024 Independent Spent Fuel Storage Installation - 2023 Annual Radiological Environmental Monitoring Program Report ML24071A1162024-05-0101 May 2024 Issuance of Amendment Nos. 244 and 232 Revise TS 3.7.8, Cooling Water (Cl) System ML24128A0882024-04-30030 April 2024 Submittal of Updated Safety Analysis Report (Usar), Revision 38 ML24089A2382024-04-29029 April 2024 Summary of Nuclear Property Insurance IR 05000282/20244012024-04-25025 April 2024 – Security Baseline Inspection Report 05000282/2024401 and 05000306/2024401 ML24100A8042024-04-24024 April 2024 – Alternative Request RR-09 for Safety Injection and Volume Control System Category C Check Valve Inservice Testing ML24114A0882024-04-23023 April 2024 Annual Report of Individual Monitoring for the Prairie Island Nuclear Generating Plant (PINGP) ML24113A1182024-04-12012 April 2024 NRC Letter Re NRC Office of Investigations Report No. 3-2023-004 ML24100A1212024-04-0909 April 2024 Submittal of Revised Pressure and Temperature Limits Report L-PI-24-012, Independent Spent Fuel Storage Installation - License Amendment Request: Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2)2024-04-0202 April 2024 Independent Spent Fuel Storage Installation - License Amendment Request: Revise Independent Spent Fuel Storage Installation (ISFSI) License Conditions 23(a), 24(A)(2), and 24(B)(2) ML24093A2832024-04-0202 April 2024 Nuclear Material Transaction Report ML24089A2402024-03-29029 March 2024 Guarantee of Payment of Deferred Premiums ML24060A1232024-03-27027 March 2024 to Request 1-RR-5-10 and 2-RR-5-10 Regarding Reactor Pressure Vessel Welds and Nozzle Welds L-PI-24-004, Independent Spent Fuel Storage Installation - Annual Effluent Report, January Through December 20232024-02-29029 February 2024 Independent Spent Fuel Storage Installation - Annual Effluent Report, January Through December 2023 IR 05000282/20230062024-02-28028 February 2024 Annual Assessment Letter for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Report 05000282/2023006 and 05000306/2023006) ML24088A1102024-02-25025 February 2024 Fairbanks Morse (Fm) Part 21 Notification Report Number 23-01 Re Asco Stainless Steel Solenoid Valves L-PI-24-009, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Unit 2, Alternative RR-09 Safety Injection System and Volume Control System Category C Check Valve Quarterly Testing2024-02-13013 February 2024 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Unit 2, Alternative RR-09 Safety Injection System and Volume Control System Category C Check Valve Quarterly Testing ML24040A1712024-02-0909 February 2024 Interim Report of a Potential Deviation or Failure to Comply Associated with Bentley Systems Incorporated Autopipe Software IR 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 2024-08-05
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML24197A2012024-07-15015 July 2024 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000282/2024004 ML24149A3712024-05-29029 May 2024 (Ping) - Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection ML24141A0452024-05-20020 May 2024 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML24116A2532024-04-25025 April 2024 Final Request for Additional Information for LAR to Revise SR 3.8.1.2 Note 3 (EPID: L- 2023-LLA-0135) ML24045A0862024-02-12012 February 2024 Final RAI for Alternative RR-09 ML23335A1152023-12-0101 December 2023 NRR E-mail Capture - Prairie Island Units 1 and 2 - Request for Additional Information LAR to Revise TS 3.7.8 Required Actions ML23248A3462023-09-0505 September 2023 NRR E-mail Capture - Request for Additional Information for Monticello Nuclear Generating Plant and Prairie Island Nuclear Generating Plant - Decommissioning Funding Status Reports ML23214A2032023-08-0202 August 2023 Request for Information for an NRC Quadrennial Comprehensive Engineering Team Inspection: Inspection Report 05000282/2024010; 05000306/2024010 ML23199A0922023-07-18018 July 2023 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000306/2023004 ML23096A3082023-04-0707 April 2023 Notification of Inspection an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23055B0562023-02-27027 February 2023 Request for Information for NRC Commercial Grade Dedication Inspection Inspection Report 05000282/2023010 and 05000306/2023010 ML23053A1432023-02-22022 February 2023 Information Request to Support Upcoming Problem Identification and Resolution (Pi&R) Inspection at Prairie Island Nuclear Generating Plant ML22166A4112022-06-15015 June 2022 NRR E-mail Capture - Request for Additional Information Prairie Island Nuclear Generating Plant, Unit 2, Alternative RR-08, PIV Leakage ML22160A6022022-06-0909 June 2022 NRR E-mail Capture - Draft Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, 24-Month Operating Cycle Amendment ML22145A4152022-05-25025 May 2022 Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection 05000282/2022402 05000306/2022402 ML22131A2652022-05-11011 May 2022 NRR E-mail Capture - Request for Additional Information Xcel Energy Amendment Request to Create a Common Eplan and EOF for Monticello and Prairie Island ML22130A5792022-05-11011 May 2022 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML21321A0452021-11-10010 November 2021 Request for Additional Information: Prairie Island 24-Month Cycle Amendment Request ML21305A0102021-10-29029 October 2021 NRR E-mail Capture - Request for Additional Information Prairie Island Cooling Water Amendment ML21252A0122021-08-30030 August 2021 NRR E-mail Capture - Request for Additional Information Amendment Request to Adopt TSTF-471 and 517-T for Prairie Island ML21147A5232021-06-0303 June 2021 Prarie Island Nuclear Generating Plant, Units 1 and 2 - Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML21131A0752021-05-10010 May 2021 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000282/2021004; 05000306/2021004 ML21099A0972021-04-0909 April 2021 Information Request to Support Upcoming Temporary Instruction 2515/194 Inspection; Inspection Report 05000282/2021012 and 05000306/2021012 ML21062A0532021-03-0202 March 2021 Information Request to Support Upcomng Problem Identification and Resolution (Pi&R) Inspection at the Prairie Island Nuclear Generating Plant ML21033A6112021-02-0101 February 2021 Request for Information for an NRC Triennial Baseline Design Bases Assurance Inspection (Team); Inspection Report 05000282/2021010 and 05000306/2021010 ML20343A1292020-12-0808 December 2020 NRR E-mail Capture - Request for Additional Information ML20192A1442020-07-0707 July 2020 NRR E-mail Capture - Request for Additional Information Prairie Island License Amendment Request to Adopt TSTF-505 ML20189A1782020-07-0606 July 2020 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000282/2020004; 05000306/2020004 ML20133K0692020-05-14014 May 2020 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML20077K6242020-04-13013 April 2020 License Amendment Request - Request for Additional Information ML20052F4102020-02-21021 February 2020 Notification of Nrc Design Bases Assurance Inspection (Programs) (05000282/202010; 05000306/202010) and Initial Request for Information ML20035F1552020-02-0404 February 2020 NRR E-mail Capture - Request for Additional Information Monticello and Prairie Island Alternative Requests to Adopt Code Cases N-786-3 and N-789-3 (Epids: L-2019-LLR-0078 and L-2019-LLR-0079) ML19233A0032019-08-14014 August 2019 NRR E-mail Capture - Request for Additional Information Prairie Island Relief Requests 1-RR-10 and 2-RR-10 ML19057A1652019-02-26026 February 2019 NRR E-mail Capture - Request for Additional Information Prairie Island 50.69 Amendment Request ML18313A0832018-11-0707 November 2018 NRR E-mail Capture - Request for Additional Information Prairie Island NFPA-805 License Condition Modification Amendment Request ML18264A1912018-09-19019 September 2018 NRC Information Request (9/19/2018); Part B Items (Onsite) IP 71111.08 - E-Mailed 09/19/18 (DRS-M.Holmberg) ML18235A2982018-08-23023 August 2018 NRR E-mail Capture - Request for Additional Information Prairie Island TSTF-425 License Amendment Request ML18169A4202018-06-25025 June 2018 Supplemental Information Needed for Acceptance of Requested Licensing Action Amendment to Modify Renewed Facility Operating License Paragraph 2.C(4)(c) ML18025C0152018-01-24024 January 2018 Request for Information for an NRC Triennial Baseline Design Bases Assurance Inspection (Team): Inspection Report 05000282/2018011; 05000306/2018011 (DRS-A.Dunlop) ML17277B3332017-10-0404 October 2017 NRR E-mail Capture - Request for Additional Information Prairie Island Special Heavy Lifting Devices LAR ML17249A9232017-09-0606 September 2017 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000282/2017004; 05000306/2017004 (Exf) ML17235A9982017-08-23023 August 2017 NRR E-mail Capture - Request for Additional Information Prairie Island EAL Scheme Change ML17221A3892017-08-0909 August 2017 NRR E-mail Capture - Request for Additional Information for Prairie Island Nuclear Generating Plant License Amendment Request Dated February 23, 2017 Emergency Response Organization ML17219A0762017-08-0707 August 2017 NRR E-mail Capture - Request for Additional Information for Prairie Island Nuclear Generating Plant License Amendment Request Dated February 23, 2017 Emergency Response Organization ML17038A5132017-02-0707 February 2017 NRR E-mail Capture - Prairie Island NFPA 805 LAR, PRA RAI 21.01 ML17018A4272017-01-18018 January 2017 NRR E-mail Capture - Request for Additional Information: Prairie Island License Amendment Request to Revise Technical Specification 3.8.7 to Remove Non-Conservative Required Action ML16326A3532016-11-18018 November 2016 NRR E-mail Capture - Draft Request for Information Related to Prairie Island NFPA-805 License Amendment ML16265A1652016-09-20020 September 2016 Notification of an NRC Triennial Heat Sink Performance Inspection and Request for Information; Inspection Report 05000282/2016004; 05000306/2016004 (Gfo) ML16189A2052016-07-0707 July 2016 Notification of NRC Inspection and Request for Information ML16113A1612016-04-21021 April 2016 Information Request to Support Upcoming Problem Identification and Resolution (Pi&R) Inspection at Prairie Island Nuclear Generating Plant, Units 1 and 2 2024-07-15
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REGUI 0.,. ! << 0 In : -e./. £' 't-") ***.... Mr. Mark A. Schimmel Site Vice President UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 June 10, 2010 Prairie Island Nuclear Generating Plant Northern States Power Company -Minnesota 1717 Wakonade Drive East Welch, MN 55089-9642 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO EXCLUDE THE DYNAMIC EFFECTS ASSOCIATED WITH CERTAIN POSTULATED PIPE RUPTURES FROM THE LICENSING BASIS BASED UPON APPLICATION OF LEAK-BE FORE-BREAK METHODOLOGY (TAC NOS. ME2976 AND ME2977)
Dear Mr. Schimmel:
By letter to the U.S.
Nuclear Regulatory Commission (NRC) dated December 22, 2009 (Agencywide Documents Access and Management System Package No. ML 100200129). Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted a request for application of a leak-before-break methodology to piping systems attached to the reactor coolant system for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on May 18, 2010, it was agreed that you would provide a response within 45 days of the date of this letter. The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.
If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely, Thomas J. Senior Mject Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and Request for Additional cc w/encl:
Distribution via REQUEST FOR ADDITIONAL INFORMATION PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306 By letter dated December 22, 2009 (Agencywide Documents Access and Management System Package Accession No. ML 100200129), the Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, requested a license amendment to allow implementation of leak-before-break (LBB) on the safety injection lines, residual heat removal lines, reactor coolant system (RCS) draindown 'line, at Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, and the pressurizer surge line at PINGP Unit 2. To complete its review, the NRC staff requests the following additional information. to the December 22, 2009 Submittal E1-1. Pages 1 and 2, Item 4 states that the safety injection line consists of a 4-inch diameter pipe and a 6-inch diameter pipe. Item 5 states that the RCS draindown line consists of a 2-inch pipe reducer and a 6-inch diameter pipe. (1) Discuss the approximate length of the 4-inch diameter and 6-inch diameter pipe segments in the safety injection line. Discuss the approximate length of the 2-inch diameter and 6-inch diameter pipe segments in the RCS draindown line. (2) Discuss whether there are pipe whip restraints installed on the 4-inch diameter portion of the safety injection piping and on the 2-inch portion of the RCS draindown line.
(3) If no pipe whip restraints are installed on the 2-inch diameter portion of the draindown line and 4-inch diameter portion of the safety injection line, and if pipe whip restraints on the 6-inch diameter portion of these two lines are removed as a result of the LBB approval, justify how the 6-inch diameter portion of these two lines is protected, should the 2-inch portion of the draindown line or the 4-inch diameter portion of the safety injection line fail in a double-end guillotine break. E1-2. Page 7. The licensee stated that when LBB was applied to the RCS loop piping in an LBB evaluation in 1986, a criterion of 1 gallon per minute (gpm) in one hour for RCS leakage was used for the leak detection system capability. However, for the current submittal, the licensee used a leakage detection limit of 0.2 gpm. The use of 0.2 gpm in the proposed LBB evaluation is an improvement in the leakage detection capability from the original licensing basis of 1 gpm. However, discuss whether the design basis for the RCS leak detection system needs to be changed in the PINGP Updated Final Safety Analysis Report and plant technical specifications via a license amendment process. If not, provide justification. Enclosure 2 to the December 22, 2009 Submittal E2-1. Page 1-1.
(1) Identify the material specification of the subject pipes and welds for LBB (e.g., desig Enclosure
-(2) Provide material properties of the pipes and welds because Table 4-1a provides only materials properties for the lower-bound shield metal arc welding process.
(3) Identify any piping analyzed for LBB in Enclosure 2 that contains Alloy 82/182 dissimilar metal welds. E2-2. Page 1-5. Table 1-1 presents 12 leak detection systems at PII\JGP with detectable leakage and response time. The licensee stated that PINGP has a very redundant leak detection system capable of detecting leakage as low as 0.1 gpm, but it is being conservative by using a leak detection capability of 0.2 gpm in the LBB analysis. The NRC staff questions the capability of the 12 detection systems and methods having the necessary redundancy and sensitivity to meet the specifications in Regulatory Guide (RG) 1.45, Revision 1. First, of the 12 detection systems and methods listed in Table 1-1, only five monitoring methods can detect a minimum leakage of 0.2 gpm or lower. Of the five monitoring methods, the operator inspection method, the daily coolant inventory method, and the sump pump operating time method would not satisfy the RG 1.45 requirement of a response time of 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The remaining two monitoring methods may be acceptable. The containment radioactive particulate monitor R-11 has an estimated response time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for a leakage rate of 0.5 gpm and it can detect a minimum of 0.1 gpm. The licensee may also take credit for the containment relative humidity monitoring which can detect a minimum leakage of 0.2 gpm with an estimated response time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for a 0.5 gpm leakage.
(1) Of the 12 leak detection systems in Table 1, confirm which leak detection systems satisfy RG 1.45. (2) On page 1-4, the licensee stated that Table 1-1 was taken from Reference 5, which is related to the PINGP coolant leakage detection system performance and was submitted to the NRC on March 31, 1976. The information in Table 1-1 is more than 30 years old. Identify the leakage detection systems and methods at Units 1 and 2 that satisfy RG 1.45, Revision 1, in terms of redundancy, reliability, and sensitivity per Standard Review Plan (SRP) Section 3.6.3.111.4.
(3) Provide the response time for the detectable leakage of 0.2 gpm because Table 1-1 presents response time based on the leakage of 0.5 gpm, 1.0 gpm and 5.0 gpm, and not 0.2 gpm. E2-3. Page 4-5, Item 3. Explain why normal operating pressure is multiplied by 1.01 for the critical flaw size calculation. E2-4. Page 4-6, Item 8.
(1) Explain why not all the nodes reported in the subject LBB evaluation have updated loading data due to the uprate conditions and why there is no stress analysis performed on the subject piping considering uprate conditions.
(2) Explain why node 1045 in the 6-inch diameter safety injection line in Table 4-10 was selected as the limiting node even though its leakage was not limiting as shown in Table 5-9. E2-5. Not used
-3E2-6. Page 4-19. Clarify why the "normal operation and [safe shutdown earthquake]
SSE" moment and leakage flaw size for node 1045 of the 6-inch diameter safety injection line in Unit 1 in Table 4-10 are not the same as the moment and leakage flaw size for node 1045 in Table 5-9. E2-7. Page 5-3, second paragraph, and Figure 5-1. For the J-R power-law representation, discuss the crack extension that was used to calculate the toughness slope dJ/da. E2-8. Pages 5-7 to 5-12. Explain why the stress corrosion cracking morphology in leak rate calculations is discussed in this report even though it does not appear that the subject pipes contain Alloy 82/182 weld material or Alloy 600 material. E2-9. Page 6-1, second paragraph, states that "...Although there was a safety injection transient in Unit 1 due to [steam generator] tube rupture in 1979, there have been no inadvertent safety injections since. This transient is therefore also considered unlikely and was not evaluated ..." The fact that a safety injection did occur in 1979 shows that the safety injection transient is a likely event and should be considered in the fatigue crack growth calculation.
Justify why the inadvertent safety injection should not be considered in the evaluation, and discuss the actions/measures that preclude the potential for having an inadvertent safety injection. E2-10. Page 6-1, second paragraph, states that "... There are no local piping system transients for the 6-inch draindown line and the 6-inch hot leg nozzles... " (1) Explain why there are no local piping system transients for the 6 inch draindown line and the 6-inch hot leg nozzles. Discuss the transients that were used for these two lines in the analysis.
(2) Discuss whether there are local piping system transients or design transients applied to other pipes in the LBB evaluation.
(3) Explain the "local" piping.system transients as opposed to the design basis transients or non-local transients.
(4) Discuss the thermal transients of the reactor coolant system draindown line and whether the thermal transients were included in the analysis. E2-11. Pages 6-2 and 6-3.
Section 6-2 discusses various stresses for crack growth evaluation. Explain why stresses due to seismic event were not discussed. to the December 22, 2009 Submittal E3-1. Page 1-1. The licensee stated that it has installed the weld overlay on the Alloy 82/182 dissimilar metal weld on the pressurizer surge line at PINGP Unit 2.
(1) Discuss whether the weld overlay is installed on the pressurizer surge line at Unit 1.
(2) Discuss inspection results of the overlaid Alloy 82/182 dissimilar metal weld(s) at Unit 2 pressurizer surge line.
(3) Provide the weld identification number for the Alloy 82/182 welds that have been weld overlaid.
E3-2. Page 4-2, Section 4.2.
(1) Discuss whether the weight of the weld overlay is included in the applied loads in the LBB evaluation. If not, provide justification.
(2) Clarify whether the loadings used in the LBB analysis are applicable for 60 years for the period of license renewal. E3-3. Page 4-4, second paragraph. The licensee stated that the thermal stratification loads are lower than safe shutdown earthquake (SSE) load as shown in Table 4-2 of Enclosure 3; thus, thermal stratification during heatup/cooldown is ignored. Justify why the thermal stratification loads in Table 4-2 are ignored because the moment in the y direction for the thermal stratification is not insignificant compared to the SSE load. If the moments for the thermal stratification in the y direction (My) are included with the SSE load in the analysis, the critical crack size and leakage crack size may be changed from the reported value in the submittal. E3-4. Page 7-1, first paragraph. The licensee stated that "...Crack growth evaluations were performed in Reference 6 to indicate that combined PWSCC and fatigue crack growth for axial and circumferential postulated flaws is within acceptable limits for [time period -
proprietary information] operating interval... " (1) Provide the acceptable limits and the starting point of the stated time period.
(2) Based on the results, the postulated flaw(s) will exceed the acceptable limits before the end of the plant license and license renewal period. Discuss how the subject pipe will be monitored to prevent flaws from exceeding the acceptable limits. Enclosure 4 to the December 22, 2009 Submittal E4-1. Page 2-1. Section 2.1 implies that stress corrosion cracking in the RCS primary loop and connecting Class 1 lines is a low probability event. The pressurizer surge line at Unit 2 contains Alloy 82/182 dissimilar metal welds, which are susceptible to primary water stress corrosion cracking (PWSCC) based on pressurized-water reactor (PWR) operating experience.
(1) In light of Alloy 82/182 welds in the surge line, explain why Section 2.1 stated that stress corrosion cracking is a low probability event and did not discuss the PWSCC issue in the pressurizer surge line. The NRC staff understands that Enclosure 3 of the submittal covers the Alloy 82/182 and PWSCC issue for the Unit 2 surge line. However, Enclosure 4 should also address the issue.
(2) Provide any prior occurrences of fatigue cracking or PWSCC in the Unit 2 pressurizer surge line. E4-2. Page 2-2, Section 2.2. Provide quantitative information about historic frequencies on water hammers in Unit 2 pressurizer surge piping. E4-3. Page 2-1, Section 2. Based on PWR operating experience, the pressurizer surge line is susceptible to thermal stratification, which is a form of thermal-induced fatigue. SRP Section 3.6.3.111.10 does not permit LBB to be applied to piping with a history of fatigue cracking
-5or failure. The licensee did not discuss thermal stratification in Section 2.0. Discuss whether thermal stratification is a concern in the pressurizer surge line at PINGP Unit 2. If not, provide technical basis. E4-4. Pages 4-1 to 4-5.
Section 4.4 states that load cases A, B, and C are normal operation conditions and 0, E, F, and G are faulted conditions. There should be 12 loading combinations of normal operation conditions and faulted conditions:
NO, NE, NF, NG, BID, B/E, B/F, BIG, C/D, C/E, C/F, and C/G. Explain why load combinations NE, NG, BID, C/D, C/E, and C/F were not shown in Table 4-3. E4-5. Page 4-5, Table 4-2.
(1) Explain how the temperatures are derived in each of the load cases in Table 4-2.
(2) Discuss whether loads caused by insurge and outsurge have been considered in the LBB evaluation. If not, provide justification. E4-6. Page 4-7, Table 4-4 provides loading for critical location, Node 1320.
(1) Discuss how the critical location, Node 1320, was selected.
(2) To aid in necessary confirmatory calculations, provide all load components (Le., moments in the x, y, and z directions and Fx) for each ASME loading category case (A, B, C and D) for Node 1320.
(3) Discuss whether the pipe loadings for the LBB evaluation include the effect of the power uprate conditions. If not, provide justification. E4-7. Page 5-3, first paragraph, states that the crack relative roughness was obtained from fatigue crack data of stainless steel samples. Discuss the source of the stainless steel samples. Discuss how the roughness value was obtained. E4-8. Page 5-3. Section 5.2.3 states that the crack opening area was estimated using the method of Reference 5-3. Discuss in detail exactly how the crack opening area was estimated and provide page numbers in Reference 5-3 which show the crack opening area calculation. E4-9. Pages 5-11 to 5-14. Discuss how the curves on these pages were constructed. E4-10. Page 6-1. The licensee did not perform a fatigue crack growth calculation for the Unit 2 pressurizer surge line. Instead, it used the results of the Unit 1 fatigue crack growth calculation to apply to the Unit 2 fatigue growth calculation. Unit 1 selected location 1 is near the reactor coolant loop nozzle and location 2 is located near the pressurizer nozzle. (1) Discuss whether the same locations in the Unit 2 surge line have the same loading as the two locations in the Unit 1 surge line.
(2) Discuss how it was determined that the two pipe locations in the Unit 1 pressurizer surge line will be the same limiting locations in the Unit 2 pressurizer surge line.
-6 (3) Discuss whether the applied loads and stresses at the Unit 1 surge line bound the applied loads at the Unit 2 surge line. E4-11. Page A-1. Cite reference(s) for the equations presented in Appendix A.
June 10, 2010 Mr. Mark A.
Schimmel Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company -Minnesota 1717 Wakonade Drive East Welch, MN 55089-9642 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO EXCLUDE THE DYNAMIC EFFECTS ASSOCIATED WITH CERTAIN POSTULATED PIPE RUPTURES FROM THE LICENSING BASIS BASED UPON APPLICATION OF LEAK-BEFORE-BREAK METHODOLOGY (TAC NOS. ME2976 AND ME2977)
Dear Mr. Schimmel:
By letter to the U.S.
Nuclear Regulatory Commission (NRC) dated December 22, 2009 (Agencywide Documents Access and Management System Package No. ML 100200129), Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted a request for application of a leak-before-break methodology to piping systems attached to the reactor coolant system for the Prairie Island Nuclear Generating Plant, Units 1 and 2.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter.
During a discussion with your staff on May 18, 2010, it was agreed that you would provide a response within 45 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.
If circumstances result in the need to revise the requested response date, please contact me at (301) 415-4037.
Sincerely, IRAJ Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
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