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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 L-P1-07-066 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 Response to Request for Additional Information on 2006 Prairie lsland Nuclear Generating Plant (PINGP) Unit 1 Steam Generator Tube Inspections In an email dated December 12, 2006, the Nuclear Regulatory Commission (NRC) Staff made the following request:
By letters dated September 1,2006 (ML062550530), and May 25,2006 (ML061450543), Nuclear Management Company (the licensee) submitted information summarizing the results of the 2006 steam generator tube inspections at Prairie lsland Unit 1. These inspections were performed during the twenty fourth refueling outage (1R24). In addition to this report, the U.S.
Nuclear Regulatory Commission staff summarized additional information concerning the 2006 steam generator tube inspections at Prairie lsland Unit 1 in a letter dated July 10, 2006 (ML061680006).
The NRC staff has reviewed the information the licensee provided and determined that additional information is required in order to complete the evaluation. The additional information being requested is enclosed. From discussions with you, the root cause evaluation is not complete. Please provide your estimated date of completion of the root cause evaluation and arrange a teleconference to discuss the staff request upon availability of the root cause evaluation report.
I . On May 18, 2006, the Nuclear Regulatory Commission (NRC) staff conducted a phone call with representatives from Prairie lsland to discuss the steam generator tube inspections during their 24th refueling outage. During the call there was a discussion of a root cause analysis that was commenced in response to finding more wear indications than expected. Please discuss the scope and results of that analysis.
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Document Control Desk Page 2 In addition, please discuss any planned corrective actions in response to the results. Also, please discuss whether there is any axial or radial pattern for the wear indications detected. If so, discuss the significance.
- 2. During the operating cycle prior to the 24th refueling outage, please discuss whether there were any chemical excursions in the steam generators (e.g.,
from a leaking condenser tube). If so, discuss the extent of the chemical excursion, the corrective action and any long term implications for the steam generators.
Nuclear Management Company, LLC (NMC) Response The NMC response was discussed during a telecon on August 23,2007. The responses provided below are based largely on that discussion. As discussed previously with the NRC Staff, the vendor root cause evaluation is not yet complete and is not required by the NRC Staff at this time.
Response to Question 1:
The NMC staff completed an apparent cause evaluation (ACE) which is summarized as follows:
The cause of the unexpected eddy current indications is a lack of industry operating experience showing tube support plate (TSP) wear in similar steam generators. Because there was no industry experience, a specific tube standard for this type of wear could not be developed. Determination of the reason there was not any industry experience is one of the tasks of the AREVA Root Cause Analysis. From preliminary evaluations of PlNGP steam generator TSP wear using the French Method of bobbin probe data analysis, the following statement is made:
"If TSP wear equivalent to that detected at PlNGP Unit 1 was present at nuclear power plants in France, it would not be detected and consequently it would not be reported applying the French data analysis rule to the bobbin probe technique."
In addition the ACE notes:
Corrective Actions:
There are several possible scenarios for the steam generator TSP wear and corrective actions will be based on the applicable scenario. These scenarios are:
This is an expected event (tube wear on TSPs) and the wear will not change in number of tubes affected or size of the individual indications after the initial operating cycle(s). This could be the reason TSP wear has
Document Control Desk Page 3 not been seen in France or other countries which have AREVA steam generators.
This is unique to the 56/19 replacement steam generators installed at PINGP, either through their design, installation, or operating conditions.
However the TSP wear will not affect a large number of tubes and continued operation will not be significantly affected. The results of the AREVA Root Cause Analysis should provide more information on this.
This is unique to the PINGP steam generators and there is a significant potential that it could affect the long term operation of the steam generators. The results of the AREVA Root Cause Analysis could lead to modifications of the Unit 1 steam generators (similar to the anti-vibration bar replacement on the original steam generators) and design changes to the potential Unit 2 replacement steam generators.
With multiple scenarios there are multiple corrective actions. Furthermore, since the corrective actions will be based in a large part on the results of the AREVA Root Cause Analysis, which is not done, all corrective actions are not presently known. Therefore the corrective actions are broken into two groups: those that are known now; and those which will be known after the root cause analysis is done.
Actions that are known now:
Update plant procedure 1H25.1, "Unit 1 Assessment of Steam Generator Tube Degradation Mechanism", with the results of the 1R24 Steam Generator Tube lnspections and the Condition Monitoring Operational Assessment.
Update the Steam Generator Strategic Plan with the results of the 1R24 Steam Generator Tube lnspections and the Condition Monitoring Operational Assessment.
Obtain TSP standards to be used for sizing TSP-tube defects.
Monitor AREVA steam generator operating experience at other plants.
Because small but significant differences exist between steam generator inspection and operation practices in the United States and elsewhere, NMC should remain in the Framatome Reactors Owners Group (FROG)
Steam Generator Technical Committee (SGTC) which provides an informational exchange between nuclear power plants with AREVA steam generators. Presently PINGP has the only 56/19 AREVA replacement steam generator in this country. Any information from other sources would be of great benefit.
Make a decision on whether to inspect the steam generators during the 1R25 outage. The results from the 1R24 Steam Generator Tube lnspections and the Condition Monitoring Operational Assessment provide a single data point; inspecting the steam generators during the next outage may provide answers to the following questions:
o Will more tubes have indications?
Document Control Desk Page 4 o Will the indications grow during future operating cycles and by how much?
Subsequent to the completion of the ACE, NMC decided to inspect the steam generators during the 1R25 outage (the next Unit 1 refueling outage). The inspection scope is provided in Table 1 below.
Table 1 IR25 Outage Steam Generator Inspection Scope INSPECTION SCOPE PROBE TYPE SIG 11 SIG 12 Full Length Bobbin 100% 100%
Rows 1 through 9 U-Bends MRPC' 0% 0%
Hot Leg Tubesheet MRPC 0% 0%
Cold Leg Tubesheet MRPC 0% 0%
Post In Situ Pressure Test MRPC 100% 100%
Tubes with indications MRPC -2002 -1 502 requiring additional inspections 1
Motorized Rotating Pancake Coil Approximate number of tubes based on 1R24 inspection results NMC has not identified an axial or radial pattern of wear indications with respect to locations within the steam generator tube bundles. The locations and numbers of indications vary between the two steam generators. The 11 Steam Generator had 57 indications on 44 tube hot legs adjacent to tube support plates from TSPs 1 through 7 with the majority adjacent to TSPs 3 and 4 and with all indications on the periphery of the bundle. The 12 Steam Generator had 7 indications on 6 tubes adjacent to TSPs which were on both the hot and cold legs with all indications on the periphery of the tube bundle.
The AREVA Root Cause analysis also considered anti-vibration bar (AVB) wear indications and concluded they were caused by a few unexpected gap configurations that have led to early initiation of wear that will stop to significantly propagate as soon as the gap configuration has stabilized. The 11 Steam Generator had 9 indications on 5 tubes adjacent to anti-vibration bars (AVBs) from AVB 3 through AVB 7. The 12 Steam Generator had 32 indications on 16 tubes adjacent to AVBs from AVB 2 through AVB 9.
The replacement steam generators have 8 Tube Support Plates and 9 Anti-vibration bars. Because the numbers and locations of indications vary significantly between the generators, the location of indications within the bundle is not thought to be significant.
Response to Question 2:
There was a small condenser leak (estimated less than eight gallons per day) that was corrected by plugging two condenser tubes. None of the EPRl secondary chemistry
Document Control Desk Page 5 guideline action levels were exceeded as a result of the leak. The indications of concern on the Unit 1 steam generators correlate strongly to mechanical wear, that is, chemistry is not expected to contribute to the indications observed during the inspection. No long term steam generator implications are anticipated as a result of the condenser tube leak.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
JA Michael D. Wadley Site Vice president, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC