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Category:Inspection Report
MONTHYEARIR 05000282/20230042024-02-0101 February 2024 Integrated Inspection Report 05000282/2023004 and 05000306/2023004 IR 07200010/20234012023-12-20020 December 2023 Independent Spent Fuel Storage Installation Security Inspection Report 07200010/2023401 IR 05000282/20234012023-12-13013 December 2023 Security Baseline Inspection Report 05000282/2023401 and 05000306/2023401 IR 05000282/20230032023-11-0808 November 2023 Integrated Inspection Report 05000282/2023003 and 05000306/2023003 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 IR 05000282/20230022023-08-0303 August 2023 Integrated Inspection Report 05000282/2023002 and 05000306/2023002 IR 05000282/20234202023-06-0101 June 2023 Security Baseline Inspection Report 05000282/2023420 and 05000306/2023420 IR 05000282/20230112023-05-16016 May 2023 Biennial Problem Identification and Resolution Inspection Report 05000282/2023011 and 05000306/2023011 IR 05000282/20230012023-05-10010 May 2023 Integrated Inspection Report 05000282/2023001 and 05000306/2023001 IR 05000282/20220062023-03-0101 March 2023 Annual Assessment Letter for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Report, Units 1 and 2 (Report 05000282/2022006 and 05000306/2022006) IR 05000282/20224042023-01-30030 January 2023 Material Control and Accounting Program Inspection Report 05000282/2022404 and 05000306/2022404 (Public) IR 05000282/20220042023-01-26026 January 2023 Integrated Inspection Report 05000282/2022004 and 05000306/2022004 IR 05000282/20224032023-01-24024 January 2023 Security Baseline Inspection Report 05000282/2022403 and 05000306/2022403 IR 05000282/20220022022-12-0808 December 2022 RE-Issue Prairie Island Nuclear Generating Plant Integrated Inspection Report 05000282/2022002 and 05000306/2022002 IR 05000282/20224022022-11-30030 November 2022 Cyber Security Inspection Report 05000282/2022402 and 05000306/2022402 IR 07200010/20222012022-11-0909 November 2022 TN-40HT, Dry Storage Cask, Inspection Report No. 07200010/2022201 IR 05000282/20220032022-11-0101 November 2022 Integrated Inspection Report 05000282/2022003 and 05000306/2022003 IR 05000282/20220112022-09-26026 September 2022 Phase 4 Post-Approval Site Inspection for License Renewal Report 05000282/2022011 and 05000306/2022011 IR 05000282/20223012022-08-31031 August 2022 NRC Initial License Examination Report 05000282/2022301; 05000306/2022301 IR 05000282/20220052022-08-29029 August 2022 Updated Inspection Plan for Prairie Island Nuclear Generating Plant (Report 05000282/2022005; 05000306/2022005) IR 05000282/20225012022-08-22022 August 2022 Emergency Preparedness Biennial Exercise Inspection Report 05000282/2022501 and 05000306/2022501 ML22227A1902022-08-16016 August 2022 Island Nuclear Generating Plant, Unit 1 - Notification of NRC Baseline Inspection and Request for Information: Inspection Report 05000282/2022004 ML22222A1732022-08-11011 August 2022 Integrated Inspection Report 05000282/2022002 and 05000306/2022002 IR 05000282/20220102022-06-24024 June 2022 Triennial Fire Protection Inspection Report 05000282/2022010 and 05000306/2022010 IR 05000282/20224012022-05-23023 May 2022 Security Baseline Inspection Report 05000282/2022401 and 05000306/2022401 ML22103A2522022-05-0404 May 2022 Review of the 2021 Steam Generator Tube Inspection Report IR 05000282/20220012022-05-0303 May 2022 Integrated Inspection Report 05000282/2022001, 05000306/2022001, and 07200010/2020001 IR 05000282/20210062022-03-0202 March 2022 Annual Assessment Letter (Report 05000282/2021006 and 05000306/2021006) IR 05000282/20210042022-02-11011 February 2022 Integrated Inspection Report 05000282/2021004 and 05000306/2021004 ML22017A0052022-01-17017 January 2022 (PINGP) 2021 Unit 2 180-Day Steam Generator Tube Inspection Report IR 05000282/20214032021-12-0202 December 2021 Security Baseline Inspection Report 05000282/2021403 and 05000306/2021403 IR 05000282/20210032021-11-12012 November 2021 Integrated Inspection Report 05000282/2021003 and 05000306/2021003 IR 05000282/20210122021-10-27027 October 2021 NRC Inspection Report 05000282/2021012 and 05000306/2021012 IR 05000282/20210052021-09-0101 September 2021 Updated Inspection Plan for Prairie Island Nuclear Generating Plant Units 1 and 2 (Report 05000282/2021005 and 05000306/2021005) IR 05000282/20215012021-08-25025 August 2021 Emergency Preparedness Biennial Exercise Inspection Report 05000282/2021501 and 05000306/2021501 IR 05000282/20210022021-08-0606 August 2021 Integrated Inspection Report 05000282/2021002 and 05000306/2021002 IR 05000282/20214022021-07-28028 July 2021 Security Baseline Inspection Report 05000282/2021402 and 05000306/2021402 IR 05000282/20210102021-06-22022 June 2021 Design Basis Assurance Inspection (Teams) Inspection Report 05000282/2021010 and 05000306/2021010 IR 05000282/20210112021-06-0303 June 2021 Biennial Problem Identification and Resolution Inspection Report 05000282/2021011 and 05000306/2021011 IR 05000282/20210012021-05-0606 May 2021 Integrated Inspection Report 05000282/2021001 and 05000306/2021001 IR 05000282/20214012021-04-21021 April 2021 Security Baseline Inspection Report 05000282/2021401 and 05000306/2021401; Independent Spent Storage Security Inspection Report 07200010/2021401 IR 05000282/20200062021-03-0404 March 2021 Annual Assessment Letter for the Prairie Island Nuclear Generating Plant (Report 05000282/2020006 and 05000306/2020006) IR 05000282/20200022021-02-16016 February 2021 Reissue - Prairie Island Nuclear Generating Plant - Integrated Inspection Report 05000282/2020002; 05000306/2020002; and 07200010/2020001 IR 05000282/20200042021-02-0404 February 2021 Integrated Inspection Report 05000282/2020004 and 05000306/2020004 ML20352A1562020-12-21021 December 2020 Review of the 2018 Steam Generator Tube Inspection Report IR 05000282/20204022020-11-12012 November 2020 Security Baseline Inspection Report 05000282/2020402 and 05000306/2020402 IR 05000282/20200032020-11-10010 November 2020 Integrated Inspection Report 05000282/2020003; 05000306/2020003; and 07200010/2020002 IR 05000282/20203012020-09-22022 September 2020 NRC Initial License Examination Report 05000282/2020301 and 05000306/2020301 IR 05000282/20200052020-09-0101 September 2020 Updated Inspection Plan for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Report 05000282/2020005 and 05000306/2020005) 2024-02-01
[Table view] Category:Letter type:L
MONTHYEARL-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 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 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 L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy 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 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 L-PI-23-010, Annual Report of Individual Monitoring2023-04-27027 April 2023 Annual Report of Individual Monitoring L-PI-23-007, Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2023-03-28028 March 2023 Supplement to Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-005, CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv)2023-03-0303 March 2023 CFR 50.55a Requests Nos. 1-RR-5-15 and 2-RR-5-15, Proposed Use of Subsequent ASME Code Edition and Addenda in Accordance with 10 CFR 50.55a(g)(4)(iv) L-PI-23-001, Day Steam Generator Tube Inspection Report2023-01-30030 January 2023 Day Steam Generator Tube Inspection Report L-PI-22-047, Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report2022-12-21021 December 2022 Resubmittal of Prairie Island Nuclear Generating Plant (PINGP) 2018 Unit 1 180-Day Steam Generator Tube Inspection Report L-PI-22-020, Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)2022-12-0202 December 2022 Application to Revise Technical Specification Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-22-040, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-10-0606 October 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts2022-09-20020 September 2022 Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts L-PI-22-032, CFR 50.46 LOCA Annual Report2022-06-16016 June 2022 CFR 50.46 LOCA Annual Report L-PI-22-033, Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles2022-06-10010 June 2022 Response to Request for Additional Information Prairie Island Nuclear Generating Plant, Units 1 and 2, Amendment to Adopt 24-Month Operating Cycles L-PI-22-003, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0707 June 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections L-PI-22-024, Supplement to Application for License Amendment to Implement 24-Month Operating Cycle2022-03-0707 March 2022 Supplement to Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-047, Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 22021-12-0707 December 2021 Response to Request for Additional Information 24-Month Cycle Amendment Prairie Island Nuclear Generating Plant, Units 1 and 2 L-PI-21-045, Response to Request for Additional Information Cooling Water System License Amendment Request2021-11-0404 November 2021 Response to Request for Additional Information Cooling Water System License Amendment Request L-PI-21-029, Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.12021-10-0707 October 2021 Exigent License Amendment Request to Revise Technical Specification 3.7.8 to Allow a One-Time Extension of the Completion Time of Required Action B.1 L-PI-21-006, License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions2021-10-0202 October 2021 License Amendment Request to Revise Technical Specification 3.3.1, Reactor Trip System (RTS) Instrumentation, to Incorporate Installed Bypass Test Capability for the Power Range RTS Functions L-PI-21-032, Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island2021-09-30030 September 2021 Response to Request for Additional Information Amendment Request to Adopt TSTF-471 and 571-T for Prairie Island L-PI-21-016, Application for License Amendment to Implement 24-Month Operating Cycle2021-08-0606 August 2021 Application for License Amendment to Implement 24-Month Operating Cycle L-PI-21-027, 2020 10 CFR 50.46 LOCA Annual Report2021-06-28028 June 2021 2020 10 CFR 50.46 LOCA Annual Report L-PI-21-023, Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report2021-05-14014 May 2021 Independent Spent Fuel Storage Installation - 2020 Annual Radiological Environmental Monitoring Program Report L-PI-21-007, Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes2021-04-19019 April 2021 Application to Revise Technical Specifications to Adopt TSTF-471, Eliminate Use of the Term Core Alterations in Actions and Notes L-PI-20-050, Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic2020-10-0707 October 2020 Request for a One-Time Exemption from 10 CFR 50, Appendix E, Biennial Emergency Preparedness Exercise Requirements, Due to COVID-19 Pandemic L-PI-20-051, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2020-09-28028 September 2020 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-20-026, Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiativ2020-09-0101 September 2020 Response to Request for Additional Information: License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4 L-PI-20-035, = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule2020-07-28028 July 2020 = Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule L-PI-20-023, Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI)2020-06-10010 June 2020 Independent Spent Fuel Storage Installation, Response to Request for Additional Information: License Amendment Request to Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI) L-PI-20-014, Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI2020-04-29029 April 2020 Supplement to License Amendment Request: Expand the Storage Capacity of the Independent Spent Fuel Storage Installation (ISFSI L-PI-20-004, License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.132020-03-30030 March 2020 License Amendment Request to Revise Technical Specifications (TS) to Remove Note I from Limiting Condition for Operating (LCO) 3.4.12 and LCO 3.4.13 L-PI-20-001, License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-12020-01-29029 January 2020 License Amendment Request to Address Issues Identified in Westinghouse Nuclear Safety Advisory Letter NSAL-09-5, Revision 1, and NSAL-15-1 L-PI-19-041, Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements2019-12-23023 December 2019 Application to Revise Technical Specifications to Adopt TSTF-547, Clarification of Rod Position Requirements L-PI-19-031, License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2019-12-16016 December 2019 License Amendment Request: Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b 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 L-PI-19-038, Submittal of Revised Pressure and Temperature Limits Report2019-09-19019 September 2019 Submittal of Revised Pressure and Temperature Limits Report L-PI-19-037, Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals2019-09-16016 September 2019 Response to Request for Additional Information: 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals L-PI-19-025, Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP)2019-08-27027 August 2019 Request to Approve Site-Specific Probabilistic Risk Assessment (PRA) Model for Flowserve N-Seal Abeyance Seal and Dynamic Testing for the Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-029, Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For...2019-08-0505 August 2019 Supplement to Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components For... L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 22019-06-13013 June 2019 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals for Prairie Island, Unit 1 and Unit 2 L-PI-19-014, Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2019-04-29029 April 2019 Response to Request for Additional Information: Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-PI-19-003, Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP)2019-02-0404 February 2019 Request for Revision to Reactor Vessel Material Surveillance Capsule Removal Schedule for Prairie Island Nuclear Generating Plant (PINGP) L-PI-19-006, Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements2019-01-29029 January 2019 Emergency License Amendment Request Regarding One-Time Extension for Technical Specification Completion Time Requirements L-PI-19-005, Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.692019-01-15015 January 2019 Online Reference Portal for NRC Review of License Amendment Request to Implement 10 CFR 50.69 L-PI-18-063, Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 8052018-12-0606 December 2018 Response to Request for Additional Information: Revise License Condition Associated with Implementation of NFPA 805 2024-01-02
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Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC SEP 0 8 2008 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie lsland Nuclear Generating Plant Unit 1 Docket 50-282 License No.
DPR-42 2008 Unit 1 180-Day Steam Generator Tube Inspection Report In accordance with Prairie lsland Nuclear Generating Plant, Unit 1 Technical Specification 5.6.7 "Steam Generator Tube lnspection Report", Nuclear Management Company submits the enclosed report of steam generator tube inspections performed during the 2008 refueling and maintenance outage on Unit 1. Summary of Commitments This letter contains no new commitments and no revisions to existing commitments. Michael D.
Wadley Site Vice President, Prairie Island d clear Generating Plant Nuclear Management Company, LLC Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC 171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone:
651.388.1 121 ENCLOSURE 1 Prairie Island Nuclear Generating Plant - Unit 1 2008 Steam Generator Tube Inspection Report In accordance with Prairie Island Nuclear Generating Plant (PINGP), Unit 1 Technical Specification 5.6.7, Nuclear Management Company (NMC) submits this report of steam generator tube inspections performed during the 2008 refueling and maintenance outage on Unit 1 (1 R25). PINGP Unit 1 has two Framatome Model 56/19 Replacement Steam Generators (RSGs) with approximately 5,600 square meters of heat transfer area utilizing tubes with 19 millimeter outside diameter. Each RSG has 4,868 thermally-treated Alloy 690 u- tubes manufactured by Sandvik which have an outside diameter of 0.750 inch and a nominal wall thickness of 0.043 inch. The tubes are configured in a square pitch of 1.0425 inches with 55 rows and 114 columns.
The tube u-bends vary in radius from 2.7000 inches for a row 1 tube to 58.9950 inches for a row 55 tube. The tubes vary in length from 738.16 inches for row 1 tubes to 923.94 inches for row 55 tubes. Row 1 through row 9 tubes were subject to stress relieving following the bending process using the thermal treatment process for an additional 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> minimum soak time. The tubes were hydraulically expanded at each end for the full depth of the tubesheet with the expansion transition being between 0.08 inches and 0.24 inches below the secondary tu besheet face. The tubesheet is low alloy steel 21.46 inches thick with alloys 82 and 182 cladding 0.375" thick for an overall thickness of 21.835 inches. The tubes are supported by eight tube support plates (TSPs) and five anti-vibration bars (AVBs) intersecting tubes between 1, 3, 5, 7 and 9 times (see Figure 1). There is one straight bar that intersects all rows at the center of each bend, two 57 degree bars that intersect rows 13 through 55 and two 14 degree bars that intersect rows 25 through 55. In addition there are 24 peripheral tubes with nine staples (one at each AVB location) that carry the entire load of the complete AVB assembly. All TSPs are constructed from Type 41 0 stainless steel. The TSPs have a minimum thickness of 1 .I81 inch and have quatrefoil-shaped holes through which the tubes pass. The AVBs are constructed from Type 405 stainless steel and are rectangular in cross section (0.5 inch by 0.3 inch).
Each RSG is equipped with a Loose Parts Trapping Systems (LPTS), which is composed of screens at the top of the downcomer and at the top of the primary (cyclone) separators. These screens (0.1 4" square mesh formed from 0.031" diameter wire), prevent foreign material from entering the steam generator tube area from the main feedwater and auxiliary feedwater systems (see Figure 1 ). Page 1 of 19
,Parts Trapping Systems ANTI-Vl8RATION Figure 1 The original Westinghouse Model 51 Steam Generators (SGs) were replaced during the 2004 refueling outage after 25.75 EFPY of operation. During the 2006 refueling outage the first inservice inspection (100%
full length bobbin) was conducted on the RSGs after accumulating the initial 1.36 EFPY of RSG operation. Based on the lack of a definitive root cause for TSP wear and only a single cycle growth rate trend for both AVB and TSP wear identified during 1 R24, the NMC conservatively elected to inspect the RSGs during 1 R25 after an additional 1.62 EFPY of RSG operation (2.98 RSG cumulative EFPY). Italicized text represents technical specification excerpts.
Each excerpt is followed by the appropriate information intended to address each specific requirement and also includes additional details based on benchmarking recent submittals and Staff requests for additional information of peer Licensees.
A legend of codes and field names is included at the end of the report. 5.6.7 Steam Generator Tube Inspection Report
- a. A report shall be submitted within 180 days after initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. Initial entry into MODE 4 occurred on March 13, 2008, dictating submittal of this report on or before September 9, 2008. The report shall include:
I. The scope of inspections performed on each SG, Table 1 and the text that follows, provides the scope of inspections performed during 1 R25.
TABLE 1 2. Active degradation found, SCOPE Full Length SupplementalO Plug Upper InternalsO Top of TubesheetO I n-bundle Inspection@
PLPO Primaw Side lns~ections - TSP wear and AVB wear were found in both SGs during 1 R25 and captured within the corrective action process. There was no significant change in either the number of new TSP and AVB indications or in the number of tubes with new TSP and AVB indications. The Operational Assessment of indication growth rates, showed that the percent through-wall per effective full power year (O/oTW/EFPY) decreased for both TSP and AVB wear in both steam generators as compared to 1 R24 results. Page 3 of 19 The scope of inspections is provided as a percentage followed by the total number of tests parenthetically where practical.
0 Supplemental MRPC@ testing (including the +pointB coil) was based on bobbin results to: 1) inspect all BLG, DNG, DNI, MBM, NQI, OXP and PDS signals for latent tube degradation, and 2) inspect all percent through wall calls to refute/confirm, characterize (axial, circumferential or volumetric) andlor measure the length of wear indications.
Notes: For clarity, only three digit codes that require supplemental MRPC@ testing and utilized during 1 R25 are included in O above. BLG is called at 2 1.0 Volt outside the tubesheets and 2 15.0 Volts inside the tubesheets, DNG is called at 2 1.0 Volt, and all the other codes above (DNI, MBM, NQI, OXP and PDS) do not have a voltage calling criteria. The 1.0 Volt DNG calling criteria was established at half the industry standardized 2 Volt calling criteria because all ding signals greater than 2 Volts were rejected in the tubing mill and we elected to establish a sample of dings to track and detect incipient degradation.
O Inspection of upper internals included the Feed Ring, J-tubes, Feedwater Ring Helix, Moisture Separators, Downcomer, LPTS and other upper bundle components per NRC Generic Letter 97-06 and Prairie Island Unit 1 56/19 Replacement Steam Generator Operation and Maintenance Manual. O Tube lane and periphery of the tube bundle inspected using Camera Transporter System. @ Random fiber-optic inspection of one out of every six columns. O Locating possible loose part (PLP) indications for investigation and possible removal based on eddy current results (not necessary).
TECHNIQUE Bobbin MRPC@ Visual Visual Visual Visual Visual SIG 11 100% (4868) 100% (277)
NIA 100% 100% -1 7% NIA ~~~~~ SIG 12 100% (4865) 100% (1 99) 100% (6) 100% NIA N/A NIA Secondary Side Inspections - The upper bundle inspection found part of the feedwater ring inspection port gasket on the Downcomer LPTS and loose bolts on the inspection ports in the 12 Steam Generator. Also there were areas of a thin layer of debris on 11 and 12 Steam Generator Downcomer LPTS. However an anticipated missing part from 11 Steam Generator Feedwater Regulating Valve (part of an elastomer ring) was not found. The top of tubesheet and in bundle inspection found no loose parts or degraded components in 11 Steam Generator. All secondary side issues were entered into the corrective action process.
- 3. Nondestructive examination techniques utilized for each degradation mechanism, Table 2 and the text that follows, provides the Electric Power Research Institute (EPRI) Examination Technique Specification Sheet (ETSS) (techniques) utilized during 1 R25 for active, potential, non-degradation and unexpected degradation.
TABLE 2 4. Location, orientation (if linear), and measured sizes (if available) of senlice induced indications, CLASSlFlCATlONO Active Active Potential Potential Tables 3, 4, 5 and 6 provide the location, orientation and measured size of each reported TSP wear indication and each reported AVB wear indication in each steam generator respectively for the two active degradation mechanisms found during 1 R25. All the tubes in these four tables were returned to service. Tables 7 and 8 provide the location, orientation and measured sizes of AVB wear indications in each steam generator respectively for tubes plugged during 1 R25. The O Active is synonymous with the term "existing" degradation that is found in the EPRI Steam Generator Integrity Assessment Guidelines. Therefore the classical definition applies (i.e., one indication equates to active). O In addition:
- 1) Bobbin ETSS's 96010.1 Rev.
7, 24013.1 Rev.
2, and 96007.1 Rev.
11 were site validated for use on non-degradation (MBMs, DNGs, PDS and cold laps), 2) Bobbin ETSS's 96005.2 Rev. 9, 96001 .I Rev. 11 and 96007.1 Rev. 11 were site validated for unexpected pitting, wastage and outside diameter stress corrosion cracking (ODSCC) degradation, 3) +pointo ETSS's 96910.1 Rev. 10 was site validated as an alternate wear sizing technique and 4)
+pointB ETSS1s 21409.1 Rev. 5, 21410.1 Rev. 6, 20510.1 Rev. 7, 20511.1 Rev. 8 and 96511.2 Rev. 16 were site validated for unexpected ODSCC and primary water stress corrosion cracking (PWSCC) degradation.
MECHANISM Wear Wear Wear Wear Page 4 of 19 LOCATION AVB TSP Staple PLP TECHNIQUEO 96004.1 Rev. 1 1 96004.1 Rev. 1 1 96004.1 Rev. 1 1 27091.2 Rev.
0 AVB wear plugging criteria for 1 R25 was lowered to greater than 10% through-wall to establish a 4.54 EFPY inspection cycle. Within Tables 3 through 8, tubes reported with multiple VOL calls at the same ROW/COL/LOCATION confirm indications of double sided AVB wear or multiple wear location sites on multiple land contact points of Quatrefoil TSPs. Conversely, single VOL calls confirm single sided wear sites at AVB and TSP locations. One tube (R55C57) in Table 4 is reported with two bobbin coil percent through wall indications (one at each TSP edge) which was confirmed as a single TSP contact point wear scar spanning the length of the TSP. A legend of fields and codes with brief explanations is provided at the end of this enclosure for clarification purposes. Page 5 of 19 TABLE 3 Steam Generator 1 I TSP Wear Page 6 of 19 TABLE 3 Page 7 of 19 TABLE 3 TABLE 3 Steam Generator 11 TSP Wear Page 9 of 19 Page 10 of 19 Page 11 of 19 Page 12 of 19 Page I3 of 19 TABLE 6 Steam Generator 12 AVB Wear 11 12 12 12 20 2 1 2 1 2 1 51 42 42 42 53 54 54 54 0.05 0.17 0.21 0.14 VOL 6 VOL VOL AV3 AV4 AV4 AV4 -0.23 -0.1 1 -0.23 -0.19 -0.05 0.20 0.27 0.18 0.43 0.46 TABLE 6 Steam Generator 12 AVB Wear Page 140f 19 Page 15 of 19 TABLE 7 Page 16 of 19 TABLE 8 Page 17 of 19
- 5. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism, Table 9 provides the number of tubes plugged during 1 R25. TABLE 9 6. Total number and percentage of tubes plugged or repaired to date, Table 10 provides the total number and percentage of tubes plugged to date.
TABLE 10 MECHANISM AVB Wear TSP Wear 7. The results of condition monitoring, including the results of tube pulls and in-situ testing, SG 11 3 0 Condition monitoring structural and leakage integrity requirements have been demonstrated for SG tube degradation observed after the second cycle of operation of RSGs at Prairie Island Unit 1 without a need for tube pulls or in-situ testing. The degradation of interest is wear scars at AVB and TSP locations. The progression of wear at the AVB locations is limiting.
A conservative operational assessment approach shows that inspection is only required after two cycles of operation without any tube plugging.
With the largest wear scar left in service at an AVB location reduced to a maximum NDE depth of 10% through-wall, inspection is only required after three cycles of operation. SG 12 3 0 SG 12 6 0.12% PLUGGING TOTAL PERCENT 8. The effective plugging percentage for all plugging and tube repairs in each SG, SG 11 3 0.06% There have been no repairs performed on these SGs; therefore the effective plugging percentage is equivalent to that reported in Table
- 10. 9. Repair method utilized and the number of tubes repaired by each repair method, and There have been no repairs performed on these SGs. 10. The results of inspections performed under Specification 5.5.8. d. 3 for all tubes that have flaws below the F* or EF* distance, and were not plugged, The report shall include: a) identification of F* and EF* tubes, and b) location and extent of degradation. Specification 5.5.8.d.3 is not applicable to Unit
- 1. Page 18 of 19 LEGEND OF FIELDS AND CODES FIELD EXPLANATION TUBE # Distinct ROWICOL combination within each Table IND # Distinct ROWICOLILOCATION combination within each Table ROW Row number of tube location COL Column number of tube location VOLTS Measured Voltage PCT Measured percent or three digit code - see below LOCATION Affected landmark - see below ELEV-FROM Measurement in inches from the centerline of the landmark to the center of the bobbin coil indication or the lower edge of the rotating coil indication ELEV-TO Measurement in inches from the centerline of the landmark to the upper edge of the rotating coil indication LENGTH Calculated Length (ELEV-FROM - ELEV-TO) FIELD CODE PERCENT BLG DNG DNI MBM NQI OXP PDS VOL 0-1 00 EXPLANATION Bulge Signal - Bobbin Coil Ding Signal - Bobbin Coil Ding with an lndication - Bobbin Coil Manufacturing Burnish Mark - Bobbin Coil Non-Quantifiable lndication - Bobbin Coil Over-Expansion Signal - Bobbin Coil Pilger Drift Signal - Bobbin Coil Volumetric lndication - MRPC@ As measured percent through wall - Bobbin Coil LOCATION TEH Tube end hot (primary face)
TSH Tube sheet hot (secondary face)
O?H ? = First through Eighth tube support plate on hot leg side AV? ? = First through Ninth anti-vibration bar O?C ? = First through Eighth tube support plate on cold leg side TSC Tube sheet cold (secondary face)
TEC Tube end cold (primary face)
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