IR 05000282/2024004
| ML25038A065 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/11/2025 |
| From: | Dariusz Szwarc NRC/RGN-III/DORS/RPB3 |
| To: | Conboy T Northern States Power Co |
| Shared Package | |
| ML24276A229 | List: |
| References | |
| IR 2024004 | |
| Download: ML25038A065 (1) | |
Text
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT - INTEGRATED INSPECTION REPORT 05000282/2024004 AND 05000306/2024004
Dear Thomas Conboy:
On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Prairie Island Nuclear Generating Plant. On January 22, 2025, the NRC inspectors discussed the results of this inspection with Bill Engler, Operations Director, and other members of your staff. The results of this inspection are documented in the enclosed report.
Three findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Prairie Island Nuclear Generating Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at Prairie Island Nuclear Generating Plant.
February 11, 2025 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Dariusz Szwarc, Chief Reactor Projects Branch 3 Division of Operating Reactor Safety Docket Nos. 05000282 and 05000306 License Nos. DPR-42 and DPR-60
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000282 and 05000306
License Numbers:
Report Numbers:
05000282/2024004 and 05000306/2024004
Enterprise Identifier:
I-2024-004-0064
Licensee:
Northern States Power Company
Facility:
Prairie Island Nuclear Generating Plant
Location:
Welch, MN
Inspection Dates:
October 1, 2024, to December 31, 2024
Inspectors:
E. Fernandez, Senior Reactor Inspector
J. Kutlesa, Senior Emergency Preparedness Inspector
J. Park, Reactor Inspector
K. Pusateri, Resident Inspector
R. Rolph, Senior Health Physicist
E. Rosario, Trans & Storage Safety Inspector
A. Shaikh, Senior Reactor Inspector
D. Tesar, Senior Resident Inspector
Approved By:
Dariusz Szwarc, Chief
Reactor Projects Branch 3
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Prairie Island Nuclear Generating Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Satisfy ASME Code Case N-722-1 Examination Requirements for Reactor Coolant System (RCS) Pressure Boundary Welds Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000282/2024004-01 Open/Closed
[H.7] -
Documentation 71111.08P The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR Part 50, Section 50.55a(g)(6)(ii)(E), Augmented ISI requirements: Reactor coolant pressure boundary visual inspections, for the licensees failure to assure the American Society of Mechanical Engineers (ASME) Code Case N-722-1 examination requirements for the Class I RCS instrument connection pressure boundary welds were satisfied. Specifically, the licensee failed to include the nozzle-to-thimble tube welds within the scope of the bottom-mounted instrumentation (BMI) visual examination activity.
Failure to Control Deviations from Regulatory and Design-Basis Licensing Requirements for Reactor Coolant System Pressure Boundary Integrity Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000306/2024004-02 Open/Closed
[H.14] -
Conservative Bias 71111.15 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, when the licensee failed to meet certain regulatory and design-basis licensing requirements for reactor coolant system (RCS) pressure boundary integrity. Specifically, the licensee failed to ensure that RCS pressure boundary integrity was maintained with redundant isolation capability when relying on RCS pressure boundary isolation air-operated valve CV-31230 (letdown isolation valve) with the other redundant isolation air-operated valve CV-31279 in a degraded condition.
Failure to Complete Required Testing Before Returning Safety-Related Component to Service Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000306/2024004-03 Open/Closed
[H.14] -
Conservative Bias 71111.15 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50,
Section 55a, when the licensee failed to complete required testing of a PINGP Unit 2 safety-related valve following repairs. Specifically, the licensee failed to complete required testing in accordance with (IAW) ASME OM Code inservice testing (IST) requirements following repairs on reactor coolant system (RCS) pressure boundary valve CV-31279.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000282,05000306/2024-001-01 LER 2024-001-01 for Prairie Island Nuclear Generating Plant, Units 1 and 2,
Auxiliary Building Special Ventilation System Inoperable during Movement of Irradiated Fuel Assemblies 71153 Closed
PLANT STATUS
Unit 1 entered the inspection period shut down for refueling outage 1R34 and remained shut down for the duration of the inspection period.
Unit 2 began the inspection period at full-rated thermal power and remained at or near full-rated thermal power except for brief power reductions for surveillances and flexible power operations until October 29, 2024. On that date the Unit 2 reactor tripped, and was maintained in Mode 3, Hot Standby. Unit 2 went critical on November 10, 2024, and reached full-rated thermal power on November 11, 2024. Unit 2 remained at or near full-rated thermal power for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)walkdown of sources while bus 25 and bus 26 were protected on October 7, 2024 (2)walkdown of D2 diesel with D1 diesel out of service (OOS) on October 29, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Fire Detection Zone (FDZ) 10, reactor building Unit 1, elevation 697' on October 9, 2024
- (2) FDZ 20, reactor building Unit 1, elevation 711'-6" on October 9, 2024
- (3) FDZ 29, reactor building Unit 1, elevation 733' on October 9, 2024
- (4) FDZ 32, reactor building Unit 1, elevation 755' on October 9, 2024
- (5) FDZ 97, D6 room while elevated risk and hot work prohibited on October 23, 2024
===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities in Unit 1 during refueling outage U1R34 for Prairie Island from September 23, 2024, to November 27, 2024.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
1. Steam Generator 11 Bottom Head to Tubesheet Cylindrical Ring Weld
Surface Examination
1. Magnetic Particle Examination of Integral Attachment Support Bracket
Weld H-5/IA
2. Liquid Penetrant Examination of Integral Attachment Spring Hanger
Weld H-1/IA Visual Examination
1. Bottom-Mounted Instrumentation (BMI) Penetrations, Instrument Connections,
and Bare Metals Welding Activities
1. Replacement of Portion of Line 10-ZX-10
Reactor Vessel Internals Inservice Inspection (ISI) Activities
1. Replacement of Baffle-Former Bolts
2. Replacement of Clevis Bolts
PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection Activities (IP Section 03.02)
The inspectors verified that the licensee conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:
PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
(1)
- Evaluation 501000083129 - Inactive Leak from SF-11-1
- Evaluation 501000080684 - Boric Acid Leak 1 Reactor CLNT Loop A Flow XMTR
- Evaluation 501000079451 - CV-31213 Diaphragm Blown (active leak)
- Evaluation 501000074133 - Leaking Fitting Downstream of VC-165-2 PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04) (1 Sample)
The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations:
(1)
1. Eddy current examinations for Unit 1 steam generators 11 and 12
2. Secondary side visual examinations for Unit 1 steam generators 11 and 12
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during the Unit 2 trip on October 29, 2024.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during Unit 2 reactor startup on November 10, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)12 residual heat removal (RHR) seal leakage evaluation on October 25, 2024 (2)bus 16 resistance measurements on October 2, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 2 Yellow risk management actions with bus 16 out of service (OOS) on October 7, 2024 (2)risk associated with bypass of 2M sudden pressure relay online on October 17, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) CV-31328 operability / past operability on October 15, 2024
- (2) VC-8-2 leaking by on October 15, 2024
- (3) D1 emergency diesel generator (EDG) slow start SP 1093 failure to come up to speed past operability review on November 4, 2024
- (4) D5 EDG degraded lube oil gasket prompt operability determination on November 18, 2024
- (5) RHR and containment spray (CS) operability following reactor coolant pump (RCP)high vibes and RHR seal failure on December 9, 2024 (6)22 turbine-driven auxiliary feedwater pump (TDAFWP) cooling water (CL) suction pipe leak prompt operability determination on December 18, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated 1R34 activities from September 20, 2024, to January 16, 2025.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (3 Samples)
- (2) PMT of MV-32025 following diagnostic testing on October 28, 2024 (3)12 component cooling water pump PMT after seal replacement (Work Order [WO] 700095444) on December 11, 2024
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) Surveillance Procedure (SP) 1295, D1 Diesel Generator 6 Month Fast Start Test, Revision 77, on November 5, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) SP 1072 local leakage rate test of penetration 5 (reactor coolant drain tank pump discharge) on October 16, 2024
71114.01 - Exercise Evaluation
Inspection Review (IP Section 02.01-02.11) (1 Sample)
- (1) The inspectors observed the biennial exercise and evaluated emergency plan Joint Information Center (JIC) performance on December 11, 2024. This exercise completes the biennial exercise and evaluated emergency plan performance conducted on June 25, 2024 and documented in NRC Prairie Island Nuclear Generating Plant - Emergency Preparedness Inspection Report 05000282/2024501 and 05000306/2024501, dated August 5, 2024 (ML24215A366).
The exercise scenario simulated an event at the Prairie Island Nuclear Generating Plant that initiated with the declaration and notification of an Alert in response to a small break loss of cooling accident in containment. The scenario conditions simulated the escalation of the significance of the event, requiring the declaration of a General Emergency and Protective Action Recommendation which was communicated to the required state and local government agencies. The JIC performed all activities in accordance with regulatory requirements and plant procedures to accurately and completely transmit approved press releases.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes.
- EPLAN-03, Prairie Island Emergency Plan Annex, Revision 0
- EPLAN-05, Prairie Island Emergency Action Level (EAL) Technical Basis Document, Revision 0
- MONTICELLO NUCLEAR GENERATING PLANT AND PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: STANDARD EMERGENCY PLAN AND CONSOLIDATED EMERGENCY OPERATIONS FACILITY (EPID L-2021-LLA-0210), 03/31/2023 This evaluation does not constitute NRC approval.
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
(1)emergency plan exercise with drill and exercise performance (DEP) on December 18,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Unit 1 containment dress out and undress/exit area during the refueling, licensee surveys potentially contaminated material prior to the material leaving the contaminated area (2)survey material leaving the radiologically controlled area (RCA) in Unit 1 access control
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following radiological work:
- (1) Unit 1 containment baffle bolt repairs (2)replacement of MV 32077
- (3) Unit 1 number 11 steam generator hand hole removal
- (4) Unit 1 clevis bolt replacement High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Unit 1 regenerative heat exchanger area
- (2) Unit 1 under vessel very high radiation area
- (3) Unit 2 volume control tank area at 100 percent power Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) Unit 1 residual heat removal pit HEPA 600719 DOP due September 11, 2025
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===
- (1) Unit 1 (September 1, 2023, through September 30, 2024)
- (2) Unit 2 (September 1, 2023, through September 30, 2024)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) October 1, 2023, through September 30, 2024 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) October 1, 2023, through September 30, 2024
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)cooling water pipe 3-CL-88 leak and N-513-3 follow-up actions on December 30, 2024
- (2) AutoPipe software error follow-up on August 13, 2024
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event report (LER):
- (1) LER 2024-001-01, for Prairie Island Nuclear Generating Plant, Units 1 and 2, Auxiliary Building Special Ventilation System Inoperable during Movement of Irradiated Fuel Assemblies (ADAMS Accession No. ML24296B194). The inspectors reviewed the LER submittal. The previous LER submittal was reviewed and the inspection conclusions associated with this LER are documented in NRC Integrated Inspection Report 05000282/2024003 AND 05000306/2024003 (ML24318C498)under Inspection Results Section 71111.15. This LER is closed.
INSPECTION RESULTS
Failure to Satisfy ASME Code Case N-722-1 Examination Requirements for Reactor Coolant System (RCS) Pressure Boundary Welds Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000282/2024004-01 Open/Closed
[H.7] -
Documentation 71111.08P The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR Part 50, Section 50.55a(g)(6)(ii)(E), Augmented ISI requirements: Reactor coolant pressure boundary visual inspections, for the licensees failure to assure the American Society of Mechanical Engineers (ASME) Code Case N-722-1 examination requirements for the Class I RCS instrument connection pressure boundary welds were satisfied. Specifically, the licensee failed to include the nozzle-to-thimble tube welds within the scope of the bottom-mounted instrumentation (BMI) visual examination activity.
Description:
Prairie Island Nuclear Generating Plant (PINGP) Unit 1 reactor pressure vessel (RPV) bottom head has
- (36) BMI nozzle penetrations, which require visual examination every other outage. These penetrations are identified as Item Number B15.80 under N-722-1. The penetration nozzles are connected to thimble tubes by welds, and these instrument connection welds require visual examination once every interval. These welds are identified as Item Number B15.100 under N-722-1.
On October 3, 2024, during the last outage (U1R34) of the current fifth interval, the inspectors performed review of the licensees ISI Program Plan. The review identified that the instrument connections identified as Item Number B15.100 were not included as part of the program plan. Subsequently, the inspectors reviewed the examination records from the prior BMI visual examinations during the current fifth interval. The scope of the previous examinations included the bottom head bare metal and the
- (36) instrument penetrations. While the exam reports documented the conditions of these components in detail for each of the instrument penetrations, no such condition documentations for the thimble tube welds were included in the exam reports.
Review of Procedure SP 1421,Reactor Vessel Bottom Head Bare Metal Visual Examination, Revision 8, revealed that the procedure specifically required the bottom head penetration interface and the head surface be inspected for leakage or boric acid deposits under the insulation. However, no such requirements existed for the thimble tube welds below the penetrations in the procedure. While it may be true that the thimble tube welds also had been looked at during the prior BMI visual examinations as these welds were located in the same area, the licensee was not able to produce evidence that the visual examination method for these welds was consistent with the method required by Code Case N-722-1.
Additionally, the licensees preparation activity for the BMI visual examination included lowering portion of the RPV bottom head insulation package down just enough to gain visual access to the bare metal surface and the instrument penetrations. Due to the configuration of the insulation and the constrained geometry of the area, many of the thimble tube welds could be fully accessed only through limited angles. Without explicit inclusion of the thimble tube welds within the scope of the examination, the inspectors were concerned that these welds could be missed from being fully examined by the method required by the code case.
Corrective Actions: The licensee entered the issues into their corrective action program. The licensee revised Procedure SP 1421 to explicitly include the thimble tube welds within the scope of the BMI visual examination activity. The licensee performed the BMI direct visual examination using the revised procedure during the regularly scheduled BMI examination.
The examination did not identify any active leakage through the thimble tube welds.
Corrective Action References: CAP 501000092756
Performance Assessment:
Performance Deficiency: The licensees failure to satisfy the examination requirements of ASME Code Case N-722-1 for the RCS pressure boundary welds was contrary to 10 CFR 50.55a(g)(6)(ii)(E) and was a performance deficiency. Specifically, the licensee did not include the nozzle-to-thimble tube welds within the scope of the BMI visual examination activity.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to include the nozzle-to-thimble tube welds within the scope of the BMI visual examination activity did not provide assurance the RCS barrier integrity was maintained.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered No to questions, After a reasonable assessment of degradation, could the finding result in exceeding the reactor coolant system (RCS) leak rate for a small LOCA (leakage in excess of normal makeup)? and After a reasonable assessment of degradation, could the finding have likely affected other systems used to mitigate LOCA (e.g., Interfacing System LOCA)? under Section A of Exhibit 1 - Initiating Events Screening Questions.
Accordingly, the finding was determined to be of very low safety significance (Green).
Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate and up-to-date documentation. Through documentation reviews and interviews with licensee staff, the inspectors determined the requirements of N-722-1 were not satisfied because the examination activities were governed by the licensees ISI program that did not require the examination of the RPV bottom head nozzle-to-thimble tube welds.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations, Part 50, Section 50.55a(g)(6)(ii)(E),
Augmented ISI requirements: Reactor coolant pressure boundary visual inspections, stated,
- (1) All licensees of pressurized water reactors must augment their inservice inspection program by implementing ASME Code Case N-722-1, subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through
- (4) of this section. The inspection requirements of ASME Code Case N-722-1 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay or stress improvement.
The licensees RPV bottom head nozzle-to-thimble tube welds were fabricated with Alloy 82 material that have not been mitigated by weld overlay or stress improvement.
ASME Code Case N-722-1 required, in part, that the additional examinations of Table 1 shall be performed for pressurized water reactor plants having partial or full penetration welds in Class 1 components fabricated with Alloy 600/82/182 material.
Table 1 of Code Case N-722-1 required, in part, that the instrument connections identified as Item No. B15.100 be visually examined once per interval.
Contrary to the above, as of October 3, 2024, the licensees inservice inspection program failed to comply with the requirements of ASME Code Case N-722-1. Specifically, the licensee failed to include the nozzle-to-thimble tube welds within the scope of the BMI visual examination activity to satisfy the requirements of N-722-1. The conditions specified in paragraphs (g)(6)(ii)(E)(2) through
- (4) of Section 50.55a did not apply.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Control Deviations from Regulatory and Design-Basis Licensing Requirements for Reactor Coolant System Pressure Boundary Integrity Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000306/2024004-02 Open/Closed
[H.14] -
Conservative Bias 71111.15 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to meet certain regulatory and design-basis licensing requirements for reactor coolant system (RCS) pressure boundary integrity. Specifically, the licensee failed to ensure that RCS pressure boundary integrity was maintained with redundant isolation capability when relying on RCS pressure boundary isolation air-operated valve CV-31230 (letdown isolation valve) with the other redundant isolation air-operated valve CV-31279 in a degraded condition.
Description:
The NRC evaluated the event and documented its initial assessment in MD 8.3 Decision Documentation for Reactive Inspection (ML24276A229). On August 6, 2024, while performing plant manipulations to recover from maintenance activities, Operations attempted to isolate chemical and volume control system (CVCS) letdown from the RCS. Upon actuation, control valve CV-31279 had dual position indication (indicating Open and Closed at the same time).
The actual position of the valve disk of CV-31279 between full open and full closed was unknown, and Operations subsequently declared the valve to be inoperable, stating that it could not be relied upon to perform its isolation safety function. Despite the degraded state of CV-31279, CVCS letdown was returned to its normal lineup with RCS pressure boundary isolation valve CV-31230 placed in the Open position. Inspectors questioned whether the regulatory and design-basis licensing requirements for redundant isolation capability of RCS pressure boundary integrity was maintained with CV-31230 in the Open position and CV-31279 in an unknown degraded state.
With respect to the design-basis licensing requirements, Section 10.2.3.3.4 of the Prairie Island Updated Safety Analysis Report (USAR) requires in part that the letdown line isolation valves located close to the reactor coolant loop are tripped closed by a low-level signal from the pressurizer level instrumentation. Redundant valves in series and redundant level channels powered by separate and independent buses are provided to ensure letdown isolation even in the event of a single active failure.
With respect to regulatory requirements, 10 CFR 50.55a(c)(1) requires that Components that are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section III of the ASME BPV Code, except as provided in paragraphs (c)(2)through
- (4) of this section. 10 CFR 50.55a(c)(2)(ii) requires in part Exceptions: Isolation capability. The component is or can be isolated from the reactor coolant system by two valves in series and Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable [emphasis added]. Further, the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) as incorporated by reference in 10 CFR 50.55a requires testing to confirm the applicable reference values (such as stroke time) following valve maintenance.
The inspectors determined that CV-31230 in the Open position with CV-31279 in an unknown degraded state did not meet the regulatory and design-basis licensing requirements for redundant isolation capability of the RCS pressure boundary integrity as specified in 10 CFR 50.55a and the USAR. With CV-31279 in a degraded condition and CV-31230 in the Open position, a credible single active failure of CV-31230 could result in a complete loss of safety function, a loss of the RCS pressure boundary, and a loss of one of three fission product barriers.
Corrective Actions: The licensee performed troubleshooting on CV-31279 on September 11, 2024, and determined that the valve appeared to be full stroking. Additionally, the licensee adjusted the actuating arm of the limit switch, and then performed stroke-time testing for data gathering purposes. This data was used as the basis for an alternative request (RR-10) submitted to the NRC on January 7, 2025, for the purpose of returning CV-31279 to operation.
Corrective Action References:
501000088577, "CV-31279 Dual Lite Indication" 501000089127, "CV-31279, U2 LTDN Line ISOL," follow up 501000089348, "CV-31279, U2 LTDN Line ISOL," follow up 501000089731, "LS Actuating Arm Out of Alignment"
Performance Assessment:
Performance Deficiency: The inspectors determined the failure to comply with regulatory and design-basis licensing requirements to maintain redundant isolation capability of the RCS pressure boundary integrity was within the ability of the licensee to foresee and correct and is therefore a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The licensee continued to operate CVCS letdown and the RCS despite not meeting the regulatory and design-basis licensing requirements to maintain redundant isolation capability of the RCS pressure boundary integrity. This issue is similar to example 12.d in IMC 0612, Appendix E.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Exhibit 3 Barrier Integrity, B Reactor Coolant System (RCS) Boundary and determined that the issue screened to Green by answering No to the screening question.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. The inspectors determined that the decision to place CV-31230 in the Open position with CV-31279 in an unknown degraded state that did not maintain redundant isolation capability of the RCS with the potential for a complete loss of safety function of the RCS pressure boundary integrity in the event of a single active failure was, in part, due to a lack of conservative bias.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part.
Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into procedures, and that deviations from such standards are controlled. In addition, the NRC regulations include specific requirements in 10 CFR 50.55a regarding redundant isolation capability of the RCS pressure boundary integrity and require implementation of the inservice testing provisions in the ASME OM Code as incorporated by reference in 10 CFR 50.55a.
Contrary to the above, from August 6, 2024, through January 7, 2025, the licensee failed to ensure that measures were established to assure that applicable regulatory requirements and the design-basis, as defined in § 50.2 and as specified in the license application, for those SSCs to which 10 CFR Part 50, Appendix B, applied were correctly translated into applicable operating procedures, and that deviations from such standards were controlled. Specifically, the licensee failed to ensure that regulatory and design-basis licensing requirements were met, and that deviations from those requirements were controlled, by allowing the continued operation of the RCS without maintaining redundant isolation capability in accordance with the single failure criterion specified in the NRC regulations and the USAR.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Complete Required Testing Before Returning Safety-Related Component to Service Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000306/2024004-03 Open/Closed
[H.14] -
Conservative Bias 71111.15 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Section 55a, when the licensee failed to complete required testing of a PINGP Unit 2 safety-related valve following repairs. Specifically, the licensee failed to complete required testing in accordance with (IAW) ASME OM Code inservice testing (IST) requirements following repairs on reactor coolant system (RCS) pressure boundary valve CV-31279.
Description:
On August 6, 2024, while performing plant manipulations on Unit 2 to recover from maintenance activities, Operations attempted to isolate chemical and volume control system (CVCS) letdown from the RCS. Upon actuation, air-operated control valve CV-31279 in PINGP Unit 2 had dual position indication (indicating Open and Closed at the same time).
The actual position of the valve disk of CV-31279 between full open and full closed was unknown, and Operations subsequently declared the valve to be both Inoperable and IST Inoperable, stating that it could not be relied upon to perform its isolation safety function. As documented in Engineering Evaluation CE 501000088577, while the valve exhibited the dual light indication outside of the applicable inservice test (i.e. SP 2162), it was apparent during the condition that the valve would not be able to be tested in the closed direction and would fail if tested. ISTC-5133(a) was applied for the valve failure to exhibit the required change of obturator position and was immediately declared inoperable.
Subsequently, on September 11, 2024, the licensee performed troubleshooting and repairs under WO 700143144-0010. It was found that the limit switch actuator arm had rotated out of alignment to engage the lower limit switch preventing it from actuating the full closed switch that turns off the open indicating light. Craft adjusted the limit switch actuating arm on the actuator stem back into alignment with the limit switch. CV-31279 was stroked as part of the work plan and timed, and the light indication remained correct.
WO 700143144-0010, Revision 05 stated:
"Added Stroke Timing after limit switch adjustment for troubleshooting data gathering only. No reconfirmation band or reference band since the valve will not be in the same testing condition as SP 2162."
WO 700143144-0010, Revision 05 also stated:
Since the plant conditions from SP 2162 cannot be recreated with Unit 2 in Mode 1 and RCS pressure at 2235 PSIG as compared to Mode 5 and the RCS depressurized, the purpose of this work plan is to stroke CV-31279 and validate local indication shows the valve closes....
A contingency is built into the work plan to allow limit switch adjustment."
As documented in Engineering Evaluation CE 501000088577 (approved on October 1, 2024):
Troubleshooting was completed under WO 700143144-0010 on September 11, 2024, and provided the following additional information:
1) The valve closes normally as expected. Change of obturator position was validated.
2) Temporary repair was implemented [emphasis added] to manually rotate the limit switch arm into position (Temp repair).
3) The valve stroke is smooth and within reference band.
a. Stroke 1: 2.93 b. Stroke 2: 2.91 4) The lights are currently operating normally post manual adjustment of the limit switch arm. [emphasis added]
On October 15, 2024, at 11:49 CDT, an Operations log entry was made indicating in part:
CV-31279 is evaluated and found to be operating acceptable per ISTC-5133(c) and is considered operationally ready to perform its function. Therefore, CV-31279 is operable and available. Personnel involved: SM [Shift Manager], AOM [Assistant Operations Manager],
Ops Director, Engineering Director, IST [Inservice Test] Engineer, AOV [Air-Operated Valve]
Engineer, Reg Affairs Manager, Plant Manager, Site Vice President The ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) is incorporated by reference into 10 CFR 50.55a, Codes and standards. ASME OM Code, Subsection ISTC, paragraph ISTC-3310, Effects of Valve Repair, Replacement, or Maintenance on Reference Values, requires in part that:
When a valve or its control system has been replaced, repaired, or has undergone maintenance1 that could affect the valves performance, a new reference value shall be determined or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service.
Note 1 in paragraph ISTC-3310 specifies in part:
Adjustment of limit switches,. are examples of maintenance that could affect valve performance parameters.
ASME OM Code, Subsection ISTC, paragraph ISTC-5130, Pneumatically Operated Valves, paragraph ISTC-5133, Stroke Test Corrective Action, paragraph
- (c) requires:
Valves declared inoperable may be repaired, replaced, or the data may be analyzed to determine the cause of the deviation and the valve shown to be operating acceptable.
ASME OM Code, Subsection ISTC, paragraph ISTC-5133(e) requires:
Before returning a repaired or replacement valve to service, a test demonstrating satisfactory operation shall be performed.
ASME OM Code, Subsection ISTA, paragraph ISTA-2000, Definitions, defines the following:
maintenance: replacement of parts, adjustments, and similar actions that do not change the design (configuration and material) of an item.
repair: the process of restoring a degraded item to its original design requirements.
Based on an evaluation of the licensees activities, the inspectors determined that the work performed on CV-31279 on September 11, 2024, was a repair including a maintenance adjustment to the limit switch as documented in the WO and the Engineering Evaluation.
Therefore, the licensee was required by the ASME OM Code, Subsection ISTC, paragraph ISTC-3310 (including its Note 1) to determine a new reference value (in this instance, stroke time) or reconfirm the previous value by an inservice test conducted before the valve was returned to service. In that the licensee declared CV-31279 inoperable upon discovery of the dual light indication, the licensee considered ASME OM Code, Subsection ISTC, paragraph ISTC-5133, to be applicable to its actions in response to the failure of CV-31279. Following the repair of CV-31279, ASME OM Code, paragraph ISTC-5133, subparagraph (e), required the licensee to perform inservice testing of CV-31279 prior to returning the valve to service. In that the licensee conducted repair (or maintenance) of the valve, the licensee was not allowed to follow the provision in ASME OM Code, paragraph ISTC-5133(c), for analysis of the valve demonstrating its proper operation without testing.
The inspectors determined that when the licensee returned CV-31279 to service without completing the required ASME OM Code testing on September 11, 2024, and subsequently declared the valve to be operable on October 15, 2024, the licensee violated the ASME OM Code requirements in paragraphs ISTC-3310 and ISTC-5133, as incorporated by reference in 10 CFR 50.55a.
Corrective Actions: The licensee performed troubleshooting on CV-31279 on September 11, 2024, and determined that the valve appeared to be full stroking. Additionally, the licensee adjusted the actuating arm of the limit switch, and then performed stroke-time testing for data gathering purposes. This data was used as the basis for an alternative request (RR-10) submitted to the NRC on January 7, 2025, for the purpose of returning CV-31279 to operation.
Corrective Action References:
501000088577, "CV-31279 Dual Lite Indication" 501000089127, "CV-31279, U2 LTDN Line ISOL," follow up 501000089348, "CV-31279, U2 LTDN Line ISOL," follow up 501000089731, "LS Actuating Arm Out of Alignment"
Performance Assessment:
Performance Deficiency: Performance Deficiency: The inspectors determined that the failure to perform the required testing was within the licensees ability to foresee and correct and was therefore a Performance Deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The licensee failed to perform required Inservice Testing prior to declaring a safety-related valve as operable and relying on it to perform its safety function of maintaining the RCS pressure boundary integrity. This issue is similar to example 2.a of IMC 0612 Appendix E.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Exhibit 3 Barrier Integrity, B Reactor Coolant System (RCS) Boundary and determined that the issue screened to Green by answering No to the screening question.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. The inspectors determined that the decision to return CV-31279 to service and declare the valve operable prior to completing the required inservice test as required by the ASME OM Code as incorporated by reference in 10 CFR 50.55a was, in part, due to a lack of conservative bias.
Enforcement:
Violation: 10 CFR 50.55a (f)(4) states, Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the in-service test requirements set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and
- (3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components."
The ASME OM Code 2004, 2006 Addenda, Subsection ISTC, paragraph ISTC-3310, Effects of Valve Repair, Replacement, or Maintenance on Reference Values, requires in part that:
When a valve or its control system has been replaced, repaired, or has undergone maintenance1 that could affect the valves performance, a new reference value shall be determined or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service. And, Note 1 in paragraph ISTC-3310 specifies in part: Adjustment of limit switches,. are examples of maintenance that could affect valve performance parameters.
The ASME OM Code 2004, 2006 Addenda, Subsection ISTC, paragraph ISTC-5130, Pneumatically Operated Valves, paragraph ISTC-5133, Stroke Test Corrective Action, paragraph
- (c) requires: Valves declared inoperable may be repaired, replaced, or the data may be analyzed to determine the cause of the deviation and the valve shown to be operating acceptably. The ASME OM Code 2004, 2006 Addenda, Subsection ISTC, paragraph ISTC-5133(e) requires: Before returning a repaired or replacement valve to service, a test demonstrating satisfactory operation shall be performed.
Contrary to the above, on September 11, 2024, and October 15, 2024, the licensee failed to meet the inservice test requirements set forth in ASME OM Code 2004, 2006 Addenda.
Specifically, the licensee failed to perform an inservice test of CV-31279, demonstrating satisfactory operation, prior to returning it to service following repairs to the limit switch performed on September 11, 2024, and declaring the valve operable on October 15, 2024, as required by ISTC-3310 and ISTC 5133(e).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Annual Follow-up of AutoPipe Software Errors 71152A Inspectors performed a follow-up review of CAP actions associated with the AutoPipe software application. Corrective actions included reviews of impacted calculations with results within the margin of error identified by the software vendor. The review determined that none of the overall outcomes of the calculations were impacted by the error. Inspectors questioned whether the calculations should be quarantined to prevent their use as an input into other calculations until the errors were corrected. Licensee engineers indicated that although there may be some benefits to doing so, this was not a requirement, and they felt that their existing processes were sufficient to prevent the uncorrected calculations from being used as inputs in future calculations. Following the discussion with the inspectors, the licensee did initiate a CAP to address the inspectors questions. The inspectors did not identify any more than minor findings or violations during this review.
Observation: Follow-up of Leak on Cooling Water Piping 71152A 71152A 3-CL-88 Inspectors selected the leak on Cooling Water (CL) piping section 3-CL-88 for follow-up. On January 10, 2020, the licensee identified a through wall leak on the CL piping supplying the 122 safeguards traveling screen. The licensee wrote a CAP document and performed an operability determination via engineering which determined the piping was operable but degraded. The license performed nondestructive evaluation (NDE) to verify the structural integrity of the piping. The licensee CAP created interim actions to perform augmented examinations of similar piping locations per the code and created an Adverse Condition Monitoring Plan (ACMP) to monitor the leak daily. Augmented exams showed thinning at two other locations and the licensee scoped them into the repair plan. The licensee replaced the degraded piping on August 25, 2020, to restore it to code compliance. Inspectors determined the corrective actions were timely and adequate to correct the condition identified. The inspectors did not identify any more than minor findings or violations during this review.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 22, 2025, the inspectors presented the integrated inspection results to Bill Engler, Operations Director, and other members of the licensee staff.
- On October 24, 2024, the inspectors presented the radiation protection baseline inspection results to Thomas Conboy, Site Vice President, and other members of the licensee staff.
- On November 27, 2024, the inspectors presented the inservice inspection results to Thomas Conboy, Site Vice President, and other members of the licensee staff.
- On December 20, 2024, the inspectors presented the emergency action level and emergency plan changes inspection results to George Medina, Emergency Preparedness Manager, and other members of the licensee staff.
- On December 20, 2024, the inspectors presented the biennial exercise and evaluated emergency plan Joint Information Center (JIC) performance inspection results to George Medina, Emergency Preparedness Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
C1.1.20.5-2
Unit 2 4.16KV System Switches and Indication Checklist
Procedures
F5 Appendix A
Appendix A - Fire Strategies
Calculations
Determination of Acceptable Replacement Baffle-Former
Bolting for Prairie Island Units 1 and 2
501000090814
Cracked Weld on Snubber Support Bracket
10/07/2024
Corrective Action
Documents
501000091130
C Shaped Indication Next to Weld 1
10/14/2024
501000091944
Protruding BFBs
11/04/2024
501000092756
BMI to Flux Thimble Tube Weld Exam
11/26/2024
Corrective Action
Documents
Resulting from
Inspection
501000092757
1R34 SP1421 Bottom Head Inspection
11/26/2024
BACC Evaluation
501000074133
Leaking Fitting Downstream of VC-165-2
06/01/2023
BACC Evaluation
501000079451
CV-31213 Diaphragm Blown (active leak)
11/28/2023
BACC Evaluation
501000080684
Boric Acid Leak 1 Reactor CLNT Loop A F XMTR
01/09/2024
Engineering
Evaluations
BACC Evaluation
501000083129
Inactive Leak from SF-11-1
03/21/2024
MRS-TRC-2417
Site Validated Inspection Techniques
SG-CDMP-24-14
Prairie Island U1R34 Steam Generator Degradation
Assessment
SG-CDMP-24-21
Prairie Island 1R34 Steam Generator Condition Monitoring
and Preliminary Operational Assessment
Miscellaneous
SG-SGMP-18-15
Prairie Island Unit 1R31 Steam Generator Condition
Monitoring and Operational Assessment
NDE Reports
20V071
Visual Examination of Bottom Head Bare Metal and BMI
10/05/2020
PI-1R34-MT-001
Magnetic Particle Examination of MS Int. Attachment
(Support Bracket - CV-31099) Weld H-5/IA
09/25/2024
PI-1R34-PT-001
Liquid Penetrant Examination of SI Int. Attachment (spring
hanger) weld H-1/IA
09/23/2024
PI-1R34-UT-001
Ultrasonic Examination of SG 11 Bottom Head to Cylindrical
Ring Weld W-1
09/26/2024
PI-1R34-UT-002
Ultrasonic Examination of SG 11 Cylindrical Ring to
Tubesheet Weld W-2
09/26/2024
NDE Reports
PT-24-VT014
Visual Examination of Bottom Head Bare Metal, BMI
Penetrations, and Instrument Connection Welds
11/26/2024
FP-PE-EPM-01
Equipment, Personnel and Material Reporting
FP-PE-NDE-200
Solvent Removable Visible Dye Penetrant Examination
FP-PE-NDE-300
Dry Magnetic Particle Examination
H10.7
Fifth Interval Inservice Inspection (ISI) Plan
Procedures
SP 1421
Reactor Vessel Bottom Head Bare Metal Visual Examination
and 9
Work Orders
700042662
Replace Section of Line 10-ZX-10
10/26/2022
1C1.3-M3
Unit 1 Shutdown to Mode 3
1C1.3-M5
Unit 1 Shutdown to Mode 5
2C1.2-M1
Unit 2 Startup to Mode 1
Procedures
2C1.2-M2
Unit 2 Startup to Mode 2
Corrective Action
Documents
Resulting from
Inspection
501000091196
BACC Actions not Scoped Appropriately
10/15/2024
PINGP 1507
RHR Pump Boric Acid Corrosion Control Leak Inspection
04/06/2023
PINGP 1507
RHR Pump Boric Acid Corrosion Control Leak Inspection
10/16/2024
Engineering
Evaluations
PINGP 1507
RHR Pump Boric Acid Corrosion Control Leak Inspection
11/18/2024
Miscellaneous
Proposed
Alternative
Submission
L-PI-24-048 Prairie Island Nuclear Generating Plant
(PINGP) Unit 1 Inservice Testing Proposed Alternative
RR-10
10/14/2024
Procedures
PE 0009
4kV Switchgear Preventive Maintenance
700126719
Sensitivity Calculation for MV-32128
Work Orders
PE-B16-2 Bus 16 Inspection/Doble
10/14/2024
Procedures
FP-OP-RSK-01
Risk Monitoring and Risk Management
Corrective Action
500000327909
CE: Evaluate Where Impacted Parts Are In
05/03/2024
501000082320
Part 21 Mini-Gen Signal Gen Defect
09/13/2024
501000090226
480V Bkr RT-A Relay Failure -P1R34
09/24/2024
501000091164
CV-31328 Was Timed Going Closed at 3.56 seconds. The
Reference Range for This Stroke Time Is 0.6 to 1.8 Seconds
and the Limiting Stroke Time Is 2.4 Seconds.
10/15/2024
501000091164
Satisfactorily Completed SP 1357, Charging System Regen
HX Outlet Valve TE per WO-000700023852-0010.
10/14/2024
Documents
501000092116
CV-31084 Exceeded Limiting Stroke Time
11/08/2024
2000015288
CV-31328 Outside Limiting Stroke Time
10/15/2024
2000024617
CV-31328 Outside Limiting Stroke Time
Engineering
Evaluations
2000027874
IST Evaluation, Component: PI:2: VC:CV-31279
09/11/2024
Issue Date:
2July2024
Fairbanks Morse Defense Service Information Letter,
Mini-Gen Signal Generator, Engine Model FM 38-1/8
MR Functional
and MSPI Failure
Evaluation
Activity #501000090226, Undervoltage Relay UVTA for
Breaker 1-52/RTA Failed Timing Test in Procedure PE 4810
with a Test Value of 60.5ms during WO 700097440.
Reference Breaker S/N 850.023-2
09/24/2024
QF0739, Approval
2000017241
Were the Troubleshooting Activities Performed in
WO 700143144-0010 Considered a Repair, and If So, how
is CV-31279 in Compliance with ASME OM Code 5133
Parts c and e?
10/16/2024
Miscellaneous
QF0739, Approval
2000017241
If the Station Going to Replace the D2 Mini-Gen Signal Gen
during Unit 1 RFO Based on Industry OE, a Known Part 21
Issue That Has Been Evaluated Under CAP 501000082320,
and Recent Failure of D1 Diesel Generator That Replaced
D1 Mini-Gen Signal Gen?
11/06/2024
500000333194
TDAFWP CL Suct Pipe Leak
Operability
Evaluations
501000092317
D5 Broken Gasket on Oil Fill Cover
11/14/2024
Procedures
FG-G-DEF-01
Definitions, Abbreviations and Acronyms
Corrective Action
Documents
501000091991
CST Minimum Temperature 70 degrees F for S/G Fill. CST
Currently at 63 degrees F.
11/05/2024
Engineering
Evaluations
608000001212
Minimum Temperature for Filling Steam Generators from
Prairie Island 1R34 Shutdown Safety Plan
Prairie Island 11 RCP Seal Repair/Replacement Dec 2024
Shutdown Safety Plan
2/04/2024
Miscellaneous
QF0739, Approval
2000017241
Documents Regarding the Decision to Allow SGs to Be
Filled Below 70F
11/13/2024
700126719
MOV Post Test Data Review Worksheet
Work Orders
MV-32128 21 RHR HX CC Inlet D70
Corrective Action
Documents
Resulting from
Inspection
501000092056
11/07/2024
X-HIAW-106-100
Cooling Water Aux Building Return A-6649
D
Drawings
XH-1-123
Waste Disposal System Flow Diagram
Engineering
Evaluations
2000028876
50.59 or 72.48 Applicability Determination and
Prescreening - CC System Quarterly Test Train B
SP 1155B
CC System Quarterly Test Train B
SP 1295
D1 Diesel Generator 6 Month Fast Start Test
ST 1155B
CC Pump Baseline Test
Procedures
ST 1155B
CC Pump Baseline Test
700061127
MV32166 1 XCESS LTDN/SL WTR RTN ISOL D70
10/18/2024
700090480
MV-32025 11 TDAFWP Suction CL Supply D70
700095444
CC Pump Seal Replacement
08/30/2024
700118668
SP 1027.5-Local Leakage Rate Test of Penetration 5 (RCDT
Pump Discharge)
700135839
MV-32078 SUMP B TO 12 RHR PMP TR B D70
10/15/2024
700140363-0160
PMT CL SYS LINE 6-CL-106 Sections P6 and P3
CC-3-2 INSP CK VLV PER D72 ATT U
10/19/2024
CC System Quarterly Test Train B
SP 1070 Rx Coolant Sys Integrity Test
Work Orders
Perform SP 1089B RTS for 12 RHR Pump
Miscellaneous
N/A
Emergency Plan Exercise Scenario Xcel Energy 12/11/2024
09/26/2024
Miscellaneous
MONTICELLO NUCLEAR GENERATING PLANT AND
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS
AND 2 - ISSUANCE OF AMENDMENTS RE: STANDARD
EMERGENCY PLAN AND CONSOLIDATED EMERGENCY
OPERATIONS FACILITY (EPID L-2021-LLA-0210)
03/31/2023
EPLAN-03
Prairie Island Emergency Plan Annex
Procedures
EPLAN-05
Prairie Island Emergency Action Level (EAL) Technical
Basis Document
71151
Miscellaneous
QF04145 NRC
and MOR Data
Collection and
Submittal Forms
Fuel Performance (PWR)
various
501000036454
Clg Wtr Thru wall drip 3-CL-88
01/10/2020
501000080246
Part 21 AutoPipe Error Notification
2/21/2023
Corrective Action
Documents
501000081004
S&L Notification, Bentley Critical Error
01/19/2024
Engineering
Evaluations
6PCR01544538
CAP Number: 501000080246
2/21/2023
Miscellaneous
Memo-Critical
Error Notification
1304731
Notice of Calculations Potentially Related to Bentley Critical
Error Report 1304731
01/19/2024