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EPID:L-2022-LLR-0012, L-PI-22-004 Prairie Island Nuclear Generating Plant, Unit 2, 10 CFR 50.55a Request No. RR-08, Request for Alternative to ISTC 3630(f) (Approved, Closed) |
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MONTHYEARL-PI-22-004, L-PI-22-004 Prairie Island Nuclear Generating Plant, Unit 2, 10 CFR 50.55a Request No. RR-08, Request for Alternative to ISTC 3630(f)2022-02-0808 February 2022 L-PI-22-004 Prairie Island Nuclear Generating Plant, Unit 2, 10 CFR 50.55a Request No. RR-08, Request for Alternative to ISTC 3630(f) Project stage: Request ML22039A3402022-02-0808 February 2022 L-PI-22-004 Prairie Island Nuclear Generating Plant, Unit 2, 10 CFR 50.55a Request No. RR-08, Request for Alternative to ISTC 3630(f) Project stage: Request ML22045A4972022-02-14014 February 2022 NRR E-mail Capture - Prairie Island Nuclear Generating Plant, Unit 2 - Acceptance of Alternative Request Alternative Related PIV Leakage Monitoring Project stage: Acceptance Review L-PI-22-034, Response to RAI, Alternative RR-08 Related to Pressure Isolation Valve Monitoring and Testing2022-07-14014 July 2022 Response to RAI, Alternative RR-08 Related to Pressure Isolation Valve Monitoring and Testing Project stage: Response to RAI ML22270A3252022-09-30030 September 2022 (Prairie Island), Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-08 Project stage: Approval 2022-02-08
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Category:Code Relief or Alternative
MONTHYEARML24100A8042024-04-24024 April 2024 – Alternative Request RR-09 for Safety Injection and Volume Control System Category C Check Valve Inservice Testing ML24060A1232024-03-27027 March 2024 To Request 1-RR-5-10 and 2-RR-5-10 Regarding Reactor Pressure Vessel Welds and Nozzle Welds ML23312A0882023-11-0808 November 2023 Acceptance of Request to Revise Alternatives 1-RR-5-10 and 2-RR-5-10 ML23130A3872023-05-18018 May 2023 Request 1-RR-5-15 and 2-RR-5-15 to Use Later Edition of ASME Section XI Code for Performance of Repair/Replacement Activities ML22270A3252022-09-30030 September 2022 (Prairie Island), Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-08 ML20276A0032020-10-0202 October 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML19282A5412019-11-0505 November 2019 Relief from the Requirements of the ASME Code 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 ML19046A1662019-02-25025 February 2019 Relief from the Requirements of the ASME Code ML18270A0152018-10-19019 October 2018 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME OM Code ML16246A2002016-10-0404 October 2016 Request for Relief 1RR-4-11 and 2-RR-4-11 Associated with Certain Inservice Inspection of Component Welds for the Fourth 10-Year Interval ML15196A2132015-08-17017 August 2015 Requests 1-RR-4-9 and 2-RR-4-9 Associated with the Fourth Ten-Year Interval for the Inservice Inspection Program ML15196A2272015-08-12012 August 2015 Requests 1-RR-5-3 and 2-RR-5-3 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program ML15196A2212015-08-12012 August 2015 Requests 1-RR-5-1 and 2-RR-5-1 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program ML15079A0032015-05-0404 May 2015 Relief Requests 1-RR-5-5 and 2-RR-5-5 for Fifth 10-Year Interval for the Inservice Inspection Program ML15079A0022015-05-0404 May 2015 Relief Requests 1-RR-5-2 and 2-RR-5-2 for Fifth 10-Year Interval for the Inservice Inspection Program ML14329A1852014-12-0505 December 2014 Relief Requests for Fifth 10-Year Inservice Testing Program Interval (Tac Nos. MF3928 and MF3929) L-PI-14-117, Withdrawal of 10 CFR 50.55a Request RR-01 Associated with the Fifth Ten-Year Interval for the Inservice Test Program (TAC Nos. MF3928 and Mf 3929)2014-11-24024 November 2014 Withdrawal of 10 CFR 50.55a Request RR-01 Associated with the Fifth Ten-Year Interval for the Inservice Test Program (TAC Nos. MF3928 and Mf 3929) ML0807905632008-05-0909 May 2008 Relief Request 1-RR-4-6 for Piping Weld Examination Coverage for the Fourth Inservice Inspection Interval ML0617200152006-07-0303 July 2006 Relief, Evaluation of Relief Request 2-RR-4-6 for Reduced Examination Volume for Class 2 RHR Heat Exchanger Shell- to-Flange Weld ML0617103642006-07-0303 July 2006 Monticello Nuclear Generating Plant, Palisades Nuclear Plant, Point Beach Nuclear Plant Units 1 and 2, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Use of ASME Code Case N-513-2 L-HU-05-028, ASME Section XI, ISI Request for Relief 21 & 222005-12-19019 December 2005 ASME Section XI, ISI Request for Relief 21 & 22 L-PI-05-038, Withdrawal of 4th Ten-Year Inspection Interval Lnservice Lnspection Program Relief Request No. 1-RR-4-12005-05-0505 May 2005 Withdrawal of 4th Ten-Year Inspection Interval Lnservice Lnspection Program Relief Request No. 1-RR-4-1 ML0509601872005-04-27027 April 2005 Evaluation of Relief Request Nos. 1-RR-4-2, 1-RR-4-3, 2-RR-4-3, 1-RR-4-4, and 2-RR-4-4 for the Fourth 10-Year Inservice Inspection Interval ML0507504282005-04-0707 April 2005 Evaluation of Relief Request to Use Code Case N-661. TAC MC3883 and TAC MC3884 ML0422200422004-10-18018 October 2004 Evaluation of Relief Request No. 16 for the Unit 2 3rd 10-year Interval Inservice Inspection Program ML0418902032004-07-27027 July 2004 Code Relief, Evaluation of Relief Request No. 20, Revision 1 - for the Third 10-Year Inservice Inspection Interval ML0411301922004-05-0303 May 2004 Relief, Limited Examinations Associated with the PINGP Unit 1, Third 10-year Inservice Inspection (ISI) Interval, MB7975 ML0409802342004-04-12012 April 2004 Prairie, Units 1 and 2, Relief Request Nos. 19 and 20 Associated with the 10-Year Interval Inservice Inspection Interval L-PI-04-044, Request for Relief No. 20, Revision 0, for Units 1 & 2, 3rd Ten Year Inservice Inspection Interval2004-03-30030 March 2004 Request for Relief No. 20, Revision 0, for Units 1 & 2, 3rd Ten Year Inservice Inspection Interval ML0401607382004-02-26026 February 2004 Prairie, Units 1 and 2, Kewaunee, Point Beach, Units 1 and 2, Palisades, Re Request for Alternatives to ASME Section XI, Appendix Viii, Supplement 10 ML0332806022004-01-13013 January 2004 Prairie, Units 1 and 2, Relief, No. 15, Evaluation of Relief Request No. 15 for the Third 10-Year Interval Inservice Inspection Program ML0323301462003-10-0202 October 2003 Relief, Re the Fourth and Successive Inservice Inspection and Inservice Testing Program Intervals ML0321300802003-08-13013 August 2003 Relief, Re Using VT-1 Visual Examinations During Third 10-Year Interval Inservice Inspection Program ML0320506712003-07-30030 July 2003 Evaluation of Relief Request No. 18 to Perform a Visual Examination in Lieu of Volumetric Examination of Reactor Vessel Nozzle Inner Radius Sections Per Code Case N-648-1, MB8363 and MB8364 ML0316402462003-06-17017 June 2003 Relief Request, Third 10-Year Interval Inservice Inspection Program ML0232905782002-11-16016 November 2002 Request for Relief No. 9 for the Unit 2 Third 10-Year Interval Inservice Inspection Program ML0230102092002-11-0707 November 2002 Relief, Third 10-Year Interval Inservice Inspection Program ML0226202392002-10-0101 October 2002 Relief, Evaluation of Relief Request Associated with the Third 10-Year Interval Inservice Inspection Program (MB5388, 5399, 5390, & 5391) ML0212904282002-06-11011 June 2002 Relief Request, Related to the First Interval Inservice Inspection Program for Metal Containment ML0216105082002-05-31031 May 2002 Request for Relief No. 12 for the Unit 2 3rd 10-Year Interval Inservice Inspection Program 2024-04-24
[Table view] Category:Safety Evaluation
MONTHYEARML24221A3622024-09-27027 September 2024 Issuance of Amendment Nos. 245 and 233 Revise Technical Specification 3.8.1, AC Sources-Operating, Surveillance Requirement 3.8.1.2, Note 3 ML24071A1162024-05-0101 May 2024 Issuance of Amendment Nos. 244 and 232 Revise TS 3.7.8, Cooling Water (Cl) System ML24100A8042024-04-24024 April 2024 – Alternative Request RR-09 for Safety Injection and Volume Control System Category C Check Valve Inservice Testing ML24060A1232024-03-27027 March 2024 To Request 1-RR-5-10 and 2-RR-5-10 Regarding Reactor Pressure Vessel Welds and Nozzle Welds ML23356A0032024-01-17017 January 2024 Issuance of Amendments Revise Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) ML23215A1672023-12-15015 December 2023 Acceptance of Requested Licensing Action Amendment Request to Revise Surveillance Requirement 3.8.1.2 Note 3 ML23115A4072023-04-26026 April 2023 Correction of License Amendment Nos. 226 and 214 Adoption of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B ML22357A1002023-03-31031 March 2023 And Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments Standard Emergency Plan and Consolidated Emergency Operations Facility ML22300A2232022-11-0101 November 2022 Issuance of Amendments 241 and 229 TSTF-577 Revised Frequencies for Steam Generator Tube Inspections ML22270A3252022-09-30030 September 2022 (Prairie Island), Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-08 ML22181A0002022-08-17017 August 2022 Issuance of Amendments Reactor Trip System Power Range Instrumentation Channels ML22166A3892022-07-28028 July 2022 Issuance of Amendments 239 and 227 24-Month Operating Cycle ML22061A2062022-04-0101 April 2022 Issuance of Amendments TSTF-471, Rev. 1, TSTF-571-T, and Administrative Changes to Technical Specification Section 5.0 ML21312A0212021-11-23023 November 2021 Issuance of Amendment Nos. 237 and 225 Inoperable Cooling Water System Supply Header ML21008A0012021-03-19019 March 2021 Issuance of Amendment Nos. 236 and 224 Low Temperature Overpressure Protection ML20346A0202021-03-15015 March 2021 Issuance of Amendment Nos. 235 and 223, Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML20356A0022021-02-0505 February 2021 Issuance of Amendment Nos. 234 and 222, Revise Technical Specifications 3.2.1 and 5.6.5 Identified in Westinghouse Nuclear Safety Advisory Letters NSAL-09-05, Revision 1, and NSAL-15-1 ML20283A3422020-11-18018 November 2020 Issuance of Amendment Nos. 233 and 221 Adoption of Technical Specifications Task Force Traveler TSTF-547, Clarification of Rod Position Requirements ML20217L1852020-10-0202 October 2020 Issuance of Amendment Nos. 232 and 220 Increase the Integrated Leak Rate Test Program Type a and Type C Test Frequency ML20276A0032020-10-0202 October 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20210M0142020-09-0808 September 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 204, 231, and 219 TSTF-529 Clarify Use and Application Rules ML20230A0512020-09-0303 September 2020 Reactor Vessel Material Surveillance Capsule Withdrawal Schedules ML20153A4012020-06-0101 June 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20134H9582020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML20135G9922020-05-29029 May 2020 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code ML19276F6842019-11-12012 November 2019 Issuance of Amendments 230 and 218 Issuance of Amendments Adoption of 10 CFR 50.69 - Risk Informed Caterization and Treatment of Structure, Systems and Components of Nuclear Power Reactors ML19232A1512019-11-0707 November 2019 Issuance of Amendments Modifying the Design Basis for Quality Classification of Certain Fuel Handling Equipment ML19140A4472019-07-30030 July 2019 Issuance of Amendments Revision to National Fire Protection Association (NFPA) Standard NFPA 805 Modifications ML19177A3802019-07-0303 July 2019 Reactor Vessel Material Surveillance Capsule Withdrawal Schedules ML19128A1332019-06-0606 June 2019 Issuance of Amendments TSTF-439 Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO ML19045A4802019-04-16016 April 2019 Issuance of Amendment Nos. 226 and 214 Adoption of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B ML19046A1662019-02-25025 February 2019 Relief from the Requirements of the ASME Code ML19029A0942019-01-29029 January 2019 Issuance of Amendment No. 213, One-Time Technical Specification Change to Extend Completion Time for EDGs D5 and D6 (Emergency Circumstances) ML18270A0152018-10-19019 October 2018 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Relief from the Requirements of the ASME OM Code ML18100A7882018-05-0101 May 2018 Issuance of Amendment Special Heavy Lifting Device Nondestructive Examination Frequency (CAC Nos. MG0072 and MG0073; EPID L-2017-LLA-0280) ML17346A3612018-03-0606 March 2018 Issuance of Amendment Nos. 224 and 211 to Adopt Changes to the Emergency Plan ML17362A2022018-03-0505 March 2018 Issuance of Amendment Concerning Revision to the Prairie Island Nuclear Generating Plant, Units 1 and 2 Emergency Plan (CAC Nos. MF9345 and MF9346; EPID L-2017-LLA-0175) ML17334A1782017-11-30030 November 2017 Issuance of Amendment Request Related to Spent Fuel Pool Criticality Technical Specification Changes (CAC Nos. MF7121 and MF7122, EPID L-2015-LLA-0002) Non-Proprietary ML17310B2392017-11-28028 November 2017 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Unit Staff Qualifications (CAC Nos. MF9545, MF9546, and MF9547; EPID L-2017-LLA-0195) ML17163A0272017-08-0808 August 2017 Issuance of Amendments Transition to NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants ML17130A7162017-06-20020 June 2017 Issuance of Amendments Technical Specification 3.8.7 Inverters-Operating ML17110A2752017-05-0404 May 2017 Safety Evaluation Regarding Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation Related to Orders EA-12-049 and EA-12-051 ML16320A0212016-11-28028 November 2016 and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Review of Changes to the Northern States Power Company Quality Assurance Topical Report ML16256A5142016-10-13013 October 2016 Issuance of Amendment One-Time Extension for Technical Specification Surveillance Requirement 3.8.4.3. DC Sources - Operating ML16246A2002016-10-0404 October 2016 Request for Relief 1RR-4-11 and 2-RR-4-11 Associated with Certain Inservice Inspection of Component Welds for the Fourth 10-Year Interval ML16133A4062016-06-16016 June 2016 Issuance of Amendment Nos. 217 and 205, Adoption of Technical Specifications Task Force TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation (CAC Nos. MF6449 and MF6450) ML15264A2092015-11-30030 November 2015 Issuance of Amendment Nos. 216 and 204 Regarding Revisions to Technical Specification 3.3.3 and Renewed Facility Operating License ML15229A1762015-08-26026 August 2015 Issuance of Amendment Nos. 215 and 203 Regarding Revision to Licensing Basis Analysis for a Waste Gas Decay Tank Rupture ML15196A2132015-08-17017 August 2015 Requests 1-RR-4-9 and 2-RR-4-9 Associated with the Fourth Ten-Year Interval for the Inservice Inspection Program ML15196A2412015-08-12012 August 2015 Requests 1-RR-5-6 and 2-RR-5-6 Associated with the Fifth Ten-Year Interval for the Inservice Inspection Program 2024-09-27
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September 30, 2022
PRAIRIE ISLAND NUCLEAR GENERATING PLANT (PRAIRIE ISLAND), UNIT 2 AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NO. RR-08 (EPID L-2022-LLR-0012)
LICENSEE INFORMATION
Recipients Name and Address: Mr. Christopher P. Domingos Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089
Licensee: Northern States Power Company
Plant Name: Prairie Island Nuclear Generating Plant (Prairie Island),
Unit 2
Docket No.: 50-306
APPLICATION INFORMATION
Submittal Date: February 8, 2022
Submittal Agencywide Documents Access and Management System (ADAMS) Accession Nos.: ML22039A340 and ML22039A341
Supplement Date: July 14, 2022
Supplement ADAMS Accession No. : ML22195A251
Alternative Provision: The applicant requested an alternative under Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(z)(2).
Applicable Code Edition and Addenda: American Society of Mechanical Engineers (ASME),
Operation and Maintenance of Nuclear Power Plants (OM) Code2004 Edition through 2006 Addenda, as incorporated by reference in 10 CFR 50.55a, for the fifth 10-year inservice testing (IST) interval at Prairie Island, Unit 2.
Applicable Inservice Inspection (ISI) or IST program interval and Interval Start/End Dates:
Prairie Island Unit 2, fifth 10-year IST program interval began on December 21, 2014, and is scheduled to end on December 20, 2024. Alternative request RR-08 only requests that the alternative apply until the refueling outage (RFO) 2023.
IST Requirement: ASME OM Code, subsection ISTA, General Requirements, item ISTA-3300, Corrective Actions, requires that corrective actions requiring repair/replacement activities shall be performed in accordance with ASME Section XI, as applicable. Other corrective actions shall be performed in accordance with the Owner's quality assurance program.
ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, paragraph (e),
Analysis of Leakage Rates, requires that leakage rate measurements shall be compared with the permissible leakage rates specified by the plant Owner for a specific valve or valve combination. If leakage rates are not specified by the Owner, the following rates shall be permissible:
(1) for water, 0.5D gal/min [gallons/minute] (12.4d ml/sec [milliliter/second]) or 5 gal/min (315 ml/sec), whichever is less, at function pressure differential
(2) for air, at function pressure differential, 7.5D standard ft 3/day (58d std.
cc/min) where:
D = nominal valve size, inch.
d = nominal valve size, cm
STC-3630(f), Corrective Action, requires that valves or valve combinations with leakage rates exceeding the valves specified by Owner per ISTC-3630(e) shall be declared inoperable and be either repaired or replaced. A retest demonstrating acceptable operation shall be performed following any required corrective action before the valve is returned to service.
Brief Description of the Proposed Alternative: In lieu of repair or replacement of the pressure isolation valve (PIV) 21 Accumulator loop A check valve 2SI-6-4, the request proposes to use an evaluation and monitoring plan for leakage past this PIV until the 2023 RFO for Prairie Island, Unit 2. The monitoring plan will have the ability to evaluate increased accumulator leakage. This alternative will demonstrate that even though PIV 2SI-6-4 has exceeded its individual IST leakage rate, the leak rate to the accumulator will be maintained by check valve SI-6-3, which is currently meeting the IST PIV leakage requirement. The request was submitted in the event of a forced outage to delay the required repair or replacement of PIV 2SI-6-4 until the next planned RFO of Prairie Island, Unit 2, in 2023 if PIV 2SI-6-4 failed its leakage test.
For additional details on the licensees request, please refer to the documents located at the ADAMS Accession Nos. identified above.
STAFF EVALUATION
PIV 2SI-6-4 at Prairie Island, Unit 2, is a check valve in series with check valve 2SI-6-3 between the 21 Accumulator and the reactor coolant system (RCS) loop a cold leg to protect the lower design pressure accumulator and associated piping from the higher RCS pressure during normal operation. PIV 2SI-6-4 is categorized as an ASME OM Code, Category A, valve because leakage past the valve is consequential to the achievement of a safety function. Check valve 2SI-6-4 is a non-technical specification (TS) PIV and provides isolation between the high-pressure RCS and low-pressure safety injection (SI) system.
PIV 2SI-6-4 is tested in accordance with the 2004 Edition through 2006 Addenda of the ASME OM Code, subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-3630. Paragraph ISTC-3630 states, in part, that Category A valves with leakage requirements not based on an Owners 10 CFR Part 50, Appendix J, program shall be tested to verify their seat leakages are within the acceptable limits. Paragraph ISTC-3630(a),
Frequency, states that tests shall be conducted at least once every 2 years. Paragraph ISTC-3630(e) requires that leakage rate measur ements shall be compared with the permissible leakage rates specified by the plant Owner for a specific valve or valve combination and provides alternative leakage rates if not specified by the Owner. Paragraph ISTC-3630(f) requires that valves or valve combinations with leakage rates exceeding the valves specified by the Owner per ISTC-3630(e) shall be declared inoperable and either repaired or replaced.
Paragraph ISTC-3630(f) also requires that a retest demonstrating acceptable operation shall be performed following any required corrective action before the valve is returned to service.
The request states that PIV 2SI-6-4 is not part of its 10 CFR Part 50, Appendix J, program. As a result, the valve leakage criteria are governed by ASME OM Code, paragraphs ISTC-3630 and ISTC-3630(a), which require PIV 2SI-6-4 to be tested biennially in order to verify that the seat leakage is within acceptable limits. The licensee established the leakage rate acceptance criteria for both 2SI-6-3 and 2SI-6-4 as five gallons per minute (gpm) using the generic guidance in paragraph ISTC-3630(e).
During the October 2021 RFO (2R32), PIV 2SI-6-4 failed the IST leakage test with a leakage rate approximated to be 5.9 gpm. Check valve 2SI-6-3 passed the IST leakage testing with a measured leakage rate of 0.53 gpm. Based on diverse indication, the licensee considered the actual leakage of 2SI-6-3 to be zero gpm. Using this information, the licensee evaluated the impact of the as-found leakage rate of PIV 2SI-6-4 on TS operability. Based on its evaluation, the licensee determined that the PIV 2SI-6-4 leakage rate did not impact the operability of the 21 SI accumulator, and restarted Unit 2 without repairing PIV 2SI-6-4 contrary to the ASME OM Code requirement in paragraph ISTC-3630(f).
In NRC Inspection Report (IR) 05000282 and 05000306/2021004, dated February 11, 2022 (ML22041B542), the NRC determined that the Prairie Island, Unit 2, licensee violated the NRC regulations in 10 CFR 50.55a(f)(4) when it failed to meet the IST requirements set forth in the ASME OM Code, as incorporated by reference in 10 CFR 50.55a, after PIV 2SI-6-4 exceeded its ASME OM Code leakage acceptance criteria. In particular, the inspectors found that the licensee failed to repair or replace PIV 2SI-6-4 prior to returning the valve to service as required in ISTC-3630(f) following the failed leakage test. In IR 2021004, the inspectors noted that the licensee entered this issue into its corrective action program and planned on repairing or replacing the valve during the next RFO.
Alternative request RR-08 proposes a forward-looking alternative to the requirements of ISTC-3630 in the event of a forced outage prior to the 2023 RFO for Prairie Island, Unit 2. The request states that repairing or replacing PIV 2SI-6-4 will require the reactor to be defueled and drained in order to establish conditions that are safe to effectuate repair, including the establishment of dose levels that meet as low as reasonably achievable principles. Moving all the fuel from the reactor to the spent fuel pool and draining the reactor is resource intensive and a complex evolution. As such, the licensee asserts that a hardship or unusual difficulty would result from compliance with ISTC-3630(f) for PIV 2SI-6-4 in the event of a forced outage before the next planned Prairie Island, Unit 2, RFO in 2023.
The licensee asserts that the proposed alternative to the requirements of ISTC-3630(f) for PIV 2SI-6-4 provides an acceptable level of quality and safety by the combination of the leak-tightness of check valve 2SI-6-3 and the implementation of an additional evaluation and monitoring plan. As part of its monitoring plan, the licensee states that each accumulator is verified to be isolated and monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The licensee established an adverse condition monitoring plan to log the level of the 21 SI accumulator once per shift using the plant surveillance procedure. The licensee states that it will further evaluate and use the corrective action process to monitor the accumulator boron concentration should the SI accumulator level rise more than 5 percent in a day. If leakage is determined to impact operability, the licensee states that the corresponding TS Required Actions will be completed, and Prairie Island, Unit 2, will be shut down in order to complete an appropriate repair or replacement activity.
In a supplemental letter dated July 14, 2022, the licensee provided the leakage history of PIV 2SI-6-4 for approximately 10 years from May 23, 2012, through October 25, 2021. The history summary shows successful leakage performance of PIV 2SI-6-4 until the measured leakage of 5.9 gpm on October 25, 2021. The licensee states that a repair activity will be completed during the Prairie Island, Unit 2, RFO in 2023 to resolve the leakage issue.
Based on the described review, the NRC staff finds that the licensee has demonstrated that a hardship would exist to meet the ASME OM Code, paragraph ISTC-3630(f), requirements to repair or replace PIV 2SI-6-4 in the event of a unplanned outage of Prairie Island, Unit 2, without a compensating increase in the level of quality and safety where monitoring does not indicate a significant leakage increase in the accumulator line prior to the next RFO in 2023.
CONCLUSION
Based on the described review, the NRC staff finds that the licensee has demonstrated that a hardship would exist to meet the ASME OM Code, paragraph ISTC-3630(f), requirements to repair or replace PIV 2SI-6-4 in the event of a unplanned outage of Prairie Island, Unit 2, without a compensating increase in the level of quality and safety where monitoring does not indicate a significant leakage increase in the accumulator line until the next RFO in 2023. As a result, the NRC staff concludes that the licensees propos ed alternative request RR-08 to the ASME OM Code requirements in ISTC-3630(f) for PIV 2SI-6-4 meets 10 CFR 50.55a(z)(2). This alternative does not apply to other isolation valves that perform leakage control functions. Therefore, the NRC staff authorizes the use of proposed alternative request RR-08 at Prairie Island, Unit 2, until end of the RFO in 2023.
All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested, and granted or authorized (as appropriate), in the subject request remain applicable.
Principal Contributor: Gurjendra Bedi, NRR Thomas Scarbrough, NRR
Date: September 30, 2022
Nancy L. Salgado, Chief Plant Licensing Branch IIII Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
cc: Listserv
ML22270A325 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EMIB/BC NRR/DORL/LPL3/BC NAME RKuntz SRohrer SBailey (KHsu for) NSalgado (JWiebe for)
DATE 9/27/2022 9/28/2022 9/13/2022 9/30/2022