IR 05000263/2025001
| ML25129A071 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/12/2025 |
| From: | Dariusz Szwarc NRC/RGN-III/DORS/RPB3 |
| To: | Hafen S Northern States Power Company, Minnesota |
| References | |
| IR 2025001 | |
| Download: ML25129A071 (1) | |
Text
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT - INTEGRATED INSPECTION REPORT 05000263/2025001
Dear Shawn Hafen:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Monticello Nuclear Generating Plant. On April 16, 2025, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as a non-cited violation (NCV)
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 Monticello 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 Monticello Nuclear Generating Plant.
May 12, 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 No. 05000263 License No. DPR-22 Enclosure:
As stated cc: Distribution via LISTSERV Signed by Szwarc, Dariusz on 05/12/25
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Monticello 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. A licensee-identified non-cited violation is documented in report section: 71111.1
List of Findings and Violations
Licensee Failure to Address Degrading Performance of Safety-Related Motor Operated Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000263/2025001-01 Open/Closed
[P.5] -
Operating Experience 71152 S A finding and associated non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI,
Corrective Actions, was self-revealed on June 28, 2024, when the division 1 outboard low pressure coolant injection (LPCI) valve, MO 2012, failed to operate in mid stroke. Specifically, the licensee failed to correct a condition adverse to quality in that the motor pinion was disengaging from the motor shaft during a declutch mechanism failure, which prevented manual valve operation.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000263/2024-002-00 LER 2024-002-00 for Monticello Nuclear Generating Plant, Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure 71153 Closed LER 05000263/2024-003-00 LER 2024-003-00 for Monticello Nuclear Generating Plant,
Unanalyzed Condition due to Appendix R Legacy Cable Routing Concern 71153 Closed
PLANT STATUS
The unit began the inspection period at rated thermal power. On January 3, 2025, power was reduced to 96 percent for control rod friction testing and returned to 100 percent. On January 17, 2025, power was reduced to 96 percent for control rod friction testing and returned to 100 percent. On January 31, 2025, power was reduced to 85 percent for control rod friction testing, rod pattern adjustment, and returned to 100 percent. On February 14, 2025, power was reduced to 87 percent for control rod friction testing and returned to 100 percent. On February 15, 2025, power was reduced to 85 percent for rod pattern adjustment and returned to 100 percent. On March 3, 2025, power was reduced to 95 percent for control rod friction testing and returned to 97 percent, limited by maximum fraction limiting critical power ratio (MFLCPR).
On March 6, 2025, power was reduced to 77 percent for a rod pattern adjustment and returned to 97 percent limited by MFLCPR. On March 12, 2025, a rod pattern adjustment was made without lowering power that increased power to 98 percent still limited by MFLCPR. On March 19, 2025, the unit SCRAMED without complications and entered Mode 3. On March 22, 2025, the unit was placed in Mode 2, a startup commenced, and the unit transitioned to Mode 1. On March 23, 2025, power was increased to 88 percent limited by MFLCPR. On March 24, 2025, a control rod pattern adjustment increased power to 98 percent limited by MFLCPR and the unit began coast down. On March 28, 2025, a control rod pattern adjustment increased power to 96 percent limited by MFLCPR and the unit began coast down. The unit continued coasting down and ended the inspection period at 94 percent limited by MFLCPR.
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 onsite 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) Core spray B on February 19, 2025
- (2) Division 1 emergency diesel generator (EDG) starting air system line up prior to surveillance activity that will make the division 2 EDG inoperable
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 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 Zone (FZ) 1-E, high pressure coolant injection (HPCI) room on January 27, 2025
- (2) FZ 1-C, reactor core isolation cooling (RCIC) room and Appendix R cable enclosure on January 27, 2025
- (3) FZ 2-H, west shutdown cooling area on January 27, 2025
- (4) FZ 1-G, control rod drive (CRD) pump room on January 27,2025
- (5) FZ 14-A, upper 4kv bus area on January 29, 2025
- (6) FZ 33, emergency filtration building - third floor, on February 4, 2025
- (7) FZ 12-A, lower 4kv bus area on February 6, 2025
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during control rod drive friction testing on March 3, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated remedial as-found simulator examination on January 30, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 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) Division 2 reactor building closed cooling water heat exchanger secondary containment function during planned maintenance
- (2) Division 1 residual heat removal (RHR) torus suction valve primary containment isolation function during planned maintenance
- (3) Work practices associated with maintenance of division 1 RHR torus suction valve
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) Division 1 EDG manual start switch repair on March 11, 2025
- (2) Expanded work scope on division 2 low pressure coolant injection outboard isolation valve on March 27, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
(1)
===501000093648, steam leak from 1-HPCI-R-6 (level switch 23-90 root valve) (2)501000093648, 1-HPCI-R-6 steam leak degradation (3)501000093983, control rod B-10 depletion discrepancy (4)501000094278, during planned preventive maintenance, division 2 control room outside air damper was found partially open when it indicated and should have been full closed (5)501000095633, MO-2013, division 2 low pressure coolant injection outboard injection valve disc staking concerns
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
=
The inspectors evaluated the following temporary or permanent modifications:
- (1) Division 1 diesel generator voltage bump testing to gather data for engineering change 601000000157, EDG voltage regulator replacement
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) (2 Samples)
- (1) Work Order (WO) 700130777-0010, RPS MG set after replacement of voltage regulator, on February 18, 2025
Surveillance Testing (IP Section 03.01) (5 Samples)
- (1) Fuel channel distortion monitoring on January 3, 2025
- (2) Reactor SCRAM functional test on January 23, 2025
- (3) Core spray B leakage test on February 19, 2025
- (4) Stack wide range gas monitor functional test on February 26, 2025
- (5) WO 700119439-0010, reactor steam dome pressure low-channel calibration and channel functional test with logic system functional test
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Standby gas treatment (SBGT) A train quarterly test on February 4, 2025
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (2 Samples)
The inspectors evaluated:
- (1) Hostile aircraft on January 30, 2025
- (2) Hostile action duck-and-cover drill on February 25,
OTHER ACTIVITIES-BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned SCRAMs per 7000 Critical Hours Sample (IP Section 02.01)===
- (1) January 1, 2024 through December 31, 2024 IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (1 Sample)
- (1) January 1, 2024 through December. 31, 2024
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)
- (1) January 1, 2024 through December 31, 2024
71152 A - 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) Operating crew failure during requalification
71152 S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in foreign material exclusion issues while operating refuel bridge that might be indicative of a more significant safety issue.
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated reactor SCRAM with reactor feed pump trip and licensees response on March 19, 2025.
Event Report (IP Section 03.02) (3 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) EN 57498 Four-hour, non-emergency report for notification to another government agency in accordance with Title 10 of the Code of Federal Regulations (10 CFR 50)50.72 (b0(2)(xi)). 450-gallon tritium spill (54,000 pico curie per liter) reported to the MN State Duty Officer.
- (2) LER 2024-003-00, Monticello Nuclear Generating Plant, Unanalyzed Condition Due to Appendix R Legacy Cable Routing Concern (ML25009A024)
The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71111.15. This LER is Closed.
- (3) LER 2024-002-00, Monticello Nuclear Generating Plant, Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure, (ML24240A166)
The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152 S. This LER is Closed.
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated unplanned loss of Internet access and licensees performance on February 26,
INSPECTION RESULTS
Licensee-Identified Non-Cited Violation 71111.15 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: 10 CFR Part 50, Appendix R, Criterion III, Specific Requirements, Section G, Fire Protection of Safe Shutdown Capability, requires, in part, cables and equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment. One of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:
- Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier;
- Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; or
- of cable and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.
Contrary to the above, on November 15, 2024, the licensee identified a legacy design failure to maintain separation between redundant trains of systems necessary to achieve and maintain hot shutdown conditions. Specifically, division 2 cables for emergency diesel generator (EDG) 12 start circuitry were routed through a division 1 area without adequate fire barrier separation, therefore not meeting any one of the three necessary conditions listed above.
The original starting logic of the division 2 EDG contained two relays that provided a start signal on lockout or loss of voltage of the 1AR auxiliary transformer. This design included cables being routed through both division 1 and division 2 fire areas. In 1984, the licensee performed an EDG modification that removed the two relays from the starting logic which removed the requirement for divisional cable separation because the cables were no longer needed to achieve or maintain safe shutdown. In 2005, the licensee identified the cables as being part of safe shutdown but were improperly labeled as division 1 in the Updated Safety Analysis Report, Appendix J (USAR-J). While running fire probabilistic risk assessment (PRA) scenarios, in which multiple failures were being modeled, the licensee identified that the routing and labeling of the division 2 starting circuit cables was incorrect.
On December 17, 2024, the licensee restored divisional separation by rerouting the cables out of division 1 EDG space.
Significance/Severity: Green. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, Step 1.5, the senior reactor analyst determined that the licensee had a PRA model capable of evaluating the risk significance of the finding based on the NRC staffs evaluation in a license amendment dated July 12, 2021 (ML21148A274). The senior reactor analyst reviewed the licensees risk evaluation based on its PRA model, including the dominant cut sets and a description of how the deficiency was modeled.
The senior reactor analyst determined that the licensee used an exposure time of 1 year and a conditional probability of a spurious operation of 0.15, which was obtained from NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE). The senior reactor analyst determined that the dominant cut sets associated with core damage frequency and large early release frequency involved a fire in the lower 4kV room that results in a loss of offsite AC power, the failure of the division 1 AC bus, and damage to the emergency diesel generator start circuitry, preventing the operation of any low-pressure injection system as well as an independent failure of the fire pump.
The senior reactor analyst determined that the licensees evaluation was acceptable and indicated a change in core damage frequency less than 1E-6 per year and a change in large early release frequency less than 1E-7 per year. Therefore, the senior reactor analyst determined the finding was of very low safety significance (Green).
Corrective Action References: Root Cause Evaluation 602000031268; CAP 501000092373, Aggressor Circuit Identified; CAP 501000092314, Safe Shutdown Cable Route Question Observation: Potential Trend in Foreign Material Exclusion Issues While Operating the Refuel Bridge 71152 A The inspectors explored a potential trend in foreign material exclusion (FME) issues associated with refuel bridge operations. On November 21, 2024, the refuel bridge operator did not ensure that a worker secured paperwork during activities on the refuel bridge.
Subsequently, a sheet of paper dropped into the refuel pool and was retrieved. The licensee entered this issue into the corrective action program as CAP 501000092602, foreign material in spent fuel pool. Corrective actions on the spot ensured the worker secured all paperwork before proceeding. On March 6, 2025, the refuel bridge snagged an extension cord powering fuel pool lighting which entangled and shattered a portion of the FME barrier surrounding the refuel pool. The licensee entered this issue into the corrective action program as CAP 501000095848, FME barrier cracked during 9015-01. The licensee changed refuel bridge operating procedures to require spotters during refuel bridge movement. Long term corrective actions were assigned to determine if refuel pool lighting power cables should be permanently hardwired. During subsequent observations of refuel bridge operations the inspectors assessed that the corrective actions effectively addressed these FME issues. No findings or violations of regulatory requirements were identified.
Observation: Operating Crew Remediation After Failing Simulator Evaluation 71152 A An operations licenses crew failure during license operator retraining occurred on January 27, 2025. Crew qualifications to stand watch were withdrawn. The inspectors evaluated the remediation plan, witnessed remediation training, and observed the crew simulator reexamination. The inspectors assessed that the remediation justified reinstatement of crew qualifications. No findings or violations of regulatory requirements were identified.
Licensee Failure to Address Degrading Performance of Safety-Related Motor Operated Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000263/2025001-01 Open/Closed
[P.5] -
Operating Experience 71152 S A finding and associated non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, was self-revealed on June 28, 2024, when the division 1 outboard low pressure coolant injection (LPCI) valve, MO 2012, failed to operate in mid stroke. Specifically, the licensee failed to correct a condition adverse to quality in that the motor pinion was disengaging from the motor shaft during a declutch mechanism failure, which prevented manual valve operation.
Description:
The NRC evaluated this condition and documented its initial assessment in Management Directive 8.3, Decision Documentation for Reactive Inspection (ML25008A104).
In January of 1993, the licensee assembled safety-related MO 2012, the division 1 outboard LPCI valve motor and actuator using an incorrect motor pinion set screw that had an oversize cup on the leading end, which prevented full set screw engagement with the shaft dimple.
This resulted in the pinion moving up the shaft as the set screw slowly cut a groove while the valve was operated. On April 4, 2021, June 7, 2023, and September 5, 2023, the licensee documented in their corrective program that the declutching lever would not engage the hand wheel to operate the valve manually. Failing to incorporate internal and external operational experience, in each instance, the licensee concluded that since the safety-related function of MO 2012 was to operate electrically, there was no urgent need to repair the valve. On June 21, 2024, the licensee electrically cycled MO 2012 to support system venting with no concerns identified. On June 28, 2024, MO 2012 failed in mid stroke while electrically opening, after which the valve would not open or close either electrically or manually.
Operational Experience (OE) exists applicable to the failure of MO 2012 on June 28, 2024. In 1989, Limitorque Valve Corporation published MU-89-1 that described the importance of properly installing the motor pinion and motor pinion set screw on Limitorque valve operators.
In 2003, Information Notice (IN) 2003-15 provided OE about how the inability to declutch and manually operate a MOV indicated that the motor pinion was working its way off the motor shaft and that failure to take corrective action could result in the unavailability of the MOV to operate either manually or electrically. Additionally, in 2017 the licensee identified that MO 2013, the division 2 outboard low pressure coolant injection valve failed to declutch and permit manual operation. The licensee corrected this condition adverse to quality (CAQ) in 2019 and noted in the work order that after disassembly they found that the motor pinion gear was not seated on the shaft. The licensee repaired the valve actuator in accordance with vendor recommendations to address and correct the motor pinion slippage.
Corrective Actions: The licensee documented the issue in the corrective action program and performed a root cause evaluation which identified that in 1993 an incorrect set screw was used to pin the motor pinion to the shaft when the MO 2012 valve actuator was assembled.
After completing the root cause evaluation, the licensee reassembled the valve actuator with the correct set screw. The licensee performed an extent of condition and identified other motor operated valves with similar vulnerabilities and entered these into the corrective action program.
Corrective Action References: Quality Issue Management Document, 501000087225, MO 2012 Failed to Cycle as Expected
Performance Assessment:
Performance Deficiency: The licensee identified but failed to promptly correct the degradation of MO 2012, a CAQ, in accordance with the requirements of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, which requires Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, on April 30, 2021, June 7, 2023, and September 5, 2023, the licensee identified but failed to promptly correct the inability to declutch and operate MO 2012 manually. This uncorrected CAQ resulted in an unintentional entry into Technical Specification LCO 3.5.1 D, two LPCI subsystems inoperable for conditions other than Condition C. or G., when on June 28, 2025, MO 2012 failed in mid stroke and could not be operated either manually or electrically.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is similar to example 3 k. of IMC 0612 App E, Examples of Minor Issues, and was determined to be more than minor because it resulted in the unavailability and inability of MO 2012 to perform its safety function.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Specifically, the inspectors answered YES to the question, Does the degraded condition represent a loss of the probabilistic risk assessment (PRA) function of one train of a multi-train technical specification (TS) system for greater than its TS allowed outage time?
and determined that a detailed risk evaluation was required.
A senior reactor analyst performed a detailed risk evaluation which characterized the issue as having very low safety significance (Green). The analyst used Version 8.2.11 of the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations and the Monticello standardized plant analysis risk model, Version TLU1, to assess the significance of the finding. Internal and external events were considered. Results from the licensees probabilistic risk assessment model were reviewed and considered best-available information to assess the significance of the finding for fire and large early release frequency (LERF). The following assumptions and factors were considered in the quantification:
- The finding caused the division 1 LPCI path, loop A, to be unavailable due to a failure to open of the outboard injection valve, MO 2012.
- The Monticello SPAR model Version 8.82 was modified (TLU1), with the assistance of Idaho National Laboratory, to reflect that MO 2012 was closed, rather than in its normally open position, by adding basic event LCI-MOV-CC-2012 - LPCI loop A injection MOV-2012 fails to open. Additionally, logic was added to reflect that the LPCI loop select function aligns to the division 2 LPCI path, loop B, by default unless a medium or large loss of coolant accident has been experienced on the B reactor coolant recirculation loop.
- The exposure period for the finding was assumed to be 8.5 days, t + repair time, for scenarios involving emergency depressurization and manual LPCI injection. The basis for this assumption was that MO 2012 was cycled successfully 7 days prior to its failure under favorable differential pressure conditions (i.e., injection path vented).
However, for scenarios involving unfavorable differential pressure conditions (i.e., injection path potentially at normal operating pressure due to inboard injection valve leakage when MO 2012 was closed) the exposure period was assumed to be 104 days, to be consistent with the length of time in the prior calendar year MO 2012 was closed rather than open. A uniform exposure time of 104 days was considered as a sensitivity for all scenarios.
- Credit was given for flexible coping strategies (FLEX)
The resultant change in core damage frequency was estimated to be less than 1E-6/year in the best-estimate and sensitivity cases. Therefore, the finding was determined to be of very low safety significance (Green). The change in LERF was determined to be of very low safety significance (Green) in the best-estimate and sensitivity case. The dominant core damage sequences for the finding were driven by internal floods and fires resulting in a plant transient with an associated feedwater overfill that resulted in the unavailability of high-pressure injection due to water in the steam supply lines, successful emergency depressurization, and failure of low-pressure injection systems ultimately resulting in core damage.
Cross-Cutting Aspect: P.5 - Operating Experience: The organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee failed to effectively evaluate and implement relevant operating experience when MO 2012 failed to declutch and operate manually on April 30, 2021, June 7, 2023, and September 5, 2023.
Enforcement:
Violation: 10 CFR Appendix B Criterion XVI, Corrective Actions, requires, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, on April 30, 2021, June 7, 2023, and September 5, 2023, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee identified but failed to promptly correct the inability to declutch and operate MO 2012 manually. This uncorrected condition adverse to quality resulted in an unintentional entry into Technical Specification LCO 3.5.1 D, two LPCI subsystems inoperable for conditions other than Condition C. or G., when on June 28, 2025, MO 2012 failed in mid stroke and could not be operated either manually or electrically.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 16, 2025, the inspectors presented the integrated inspection results to S. Hafen, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
255-03-IIC-2
Core Spray System Leakage Test - Loop B
Procedures
2154-28
Diesel Generator Air Start System Prestart Valve Checklist
STRATEGY
A.3-01-C
Fire Zone 1-C,
RCIC Room and Appendix R Cable Enclosure
STRATEGY
A.3-01-E
Fire Zone 1-E,
HPCI Room - Reactor Building Elevation 896'
STRATEGY
A.3-01-G
Fire Zone 1-G,
CRD Pump Room
STRATEGY
A.3-02-H
Fire Zone 2-H,
West Shutdown Cooling Area
STRATEGY
A.3-12-A
Fire Zone 12-A, Lower 4 kV Bus Area (13 & 15)
STRATEGY
A.3-14-A
Fire Zone 14-A, Upper 4kV Bus Area (14 & 16)
Fire Plans
STRATEGY
A.3-33
Fire Zone 33, EFT Building Third Floor
Corrective Action
Documents
501000095729
Control Rods with High Cell Friction
03-03-2025
Miscellaneous
RQ SS-178
RPV Flood
2300
Reactivity Adjustment
Procedures
8397
Fuel Channel Distortion Monitoring
4131-PM
RBCCW Heat Exchanger Inspection and Cleaning
4901-04-PM
Torque Switch Adjustment Procedure for Rotork Valve
Operators
8136-04
Secondary Containment Penetration Work Control Checklist
Procedures
FP-MA-COM-01
Conduct of Maintenance
501000096012
DG1/CS Stuck in Pull to Lock
03-10-2025
501000096557
MO 2013 Work Missing 8154 Guidance
03-26-2025
501000096578
High Torque Diagnostic Test
03-27-2025
501000096586
MO 2013 As-found Over Torque
03-26-2025
Corrective Action
Documents
501000096593
Teledyne Test Probe Inoperable
03-26-2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
501000093648
Steam Leak from 1-HPIC-R-6
01-01-2025
501000093983
Control Rod B-10 Depletion Discrepancy
01-12-2025
501000093984
1-HPCI-R-6 Steam Leak Degradation
01-13-2025
501000094278
Damper Found Not Closed
01-22-2025
501000095331
Battery Room Differential Pressure
2-19-2025
501000095633
MO-2013 Staking Concerns
2-27-2025
Corrective Action
Documents
501000095633
MO 2013 Staking Concerns
2-27-2025
Engineering
Changes
601000000157
Emergency Diesel Generator Voltage Regulator
Replacement
Work Orders
700077056
EDG Bump Test
501000095601
Additional Information SPDS QIM
2-26-2025
501000096012
DG1/CS Stuck in Pull to Lock
03-10-2025
Corrective Action
Documents
501000096052
Minor EDG Drawing Errors
03-12-2025
Corrective Action
Documents
Resulting from
Inspection
501000096039
Trend in Control Switch Issues
03-12-2025
Drawings
NX-9216
Physical Scheme and Field Connections
Model #1999 11 EDG
0010
Reactor Scram Functional Test
0162-A
Stack Wide Range Gas Monitor Functional Test
253-01
SBGT A Train Quarterly Test
255-03-IIC-2
Core Spray System Leakage Test - Loop B
8397
Fuel Channel Distortion Monitoring
2300
Reactivity Maneuvering Steps
Procedures
Reactor Steam Dome Pressure Low-Channel Calibration
and Channel Functional Test with LSFT
700130777-0010
Replace RPS MG Set Voltage Regulator
Work Orders
700152341-0020
Emergency Diesel Generator Control
501000095535
Security on Unlocked Offices
2-25-2025
Corrective Action
Documents
501000095553
EP Drill - Duck-and-Cover Observations
2-25-2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
501000095548
EP Drill - Duck-and-Cover Meaning
2-25-2025
Miscellaneous
RPV Flood
5790-301-05
Emergency Preparedness Instructions
Procedures
A.2-110
Response to a Security Threat or Hostile Action
71151
Miscellaneous
QF0445
Monthly Operating Report - Generation Occurrences,
January 2024 Through December 2024
Miscellaneous
RPV Flood
71152 A
Procedures
FP-T-SAT-60
Systematic Approach to Training
MOVDS-MT 1
RHR MO-2012
Thrust and Torque Calculation MO-2012 (MNGP-1) - AC
Motor Operated GL96-05 Globe Valve
A
Calculations
PRA-CALC-
MT-24-002
SDP Calculations for MO-2012 Failure
501000051309
MO-2012 Manual Operator Doesnt Work
04-30-2021
501000076566
MO-2012 Declutch Lever Not Disengaging
09-05-2023
501000087225
MO-2012 Failed to Cycle as Expected
06-28-2024
501000088600
MO-2012 Motor Pinion May Not be Secured
08-05-2024
501000088681
MO-2012 Set Screw
08-08-2024
501000088690
WO 700141731 MO-2012 Issue
08-07-2024
501000089688
MO-2013 Motor Pinion Concern
09-10-2024
501000092602
FM in Spent Fuel Pool
11-21-2024
Corrective Action
Documents
501000095848
FME Barrier Cracked During 9015-01
03-06-2025
Corrective Action
Documents
Resulting from
Inspection
501000094127
MO-2012 Past Operability
01-16-2025
Drawings
NH-36247
P&ID Residual Heat Removal System
2145
RHR System Discharge Venting
C.4-C
Shutdown Outside Control Room
Procedures
C.5-3203
Use of Alternate Injection Systems for RPV Makeup
71152 S
Work Orders
700021302-0010
MECH-MO-2013, Repair Declutch Lever
04-22-2019
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
700134237-0010
255-04-IA-1-1 RHR Loop A Quarterly PMP/VLV
06-05-2024
700137902-0010
OSP-RHR-0556 LPCI Discharge Venting
06-28-2024
700141731-0010
MO-2012, Secure Motor Pinion
08-07-2024
501000095577
MNGP Internet Outage February 26 @ 0858
2-26-2025
501000096256
Reactor SCRAM No. 143 11 RFP Trip
03-19-2025
501000096334
Discharge Canal Temp ROC Limit Exceeded
03-19-2025
501000096376
Oil leak on Shaft Oil Pump P-2A
03-22-2025
501000096424
P-2A Tripped on Low Oil Pressure
03-22-2025
501000096428
RFP Foot Valve Replacement
03-23-2025
501000096454
Oil Pump Coupling Damaged
03-23-2025
501000096463
FW-67-1 11 RFP Disch CKV Back Leakage
03-23-2025
Corrective Action
Documents
501000096471
RFP Discharge Chk Potential Stuck Open
03-23-2025
Engineering
Evaluations
CWT Air Binding White Paper
2000033430
Troubleshooting Plan
03-19-2025
Event Notification Worksheet
01-16-2025
Miscellaneous
Event Notification Worksheet
03-19-2025
2165
SCRAM Report
2167
Plant Startup
111
2167-01
Startup Checklist Transition from Mode 3 to Mode 2
2167-05
Startup Checklist Transition from Mode 2 to Mode 1
Procedures
204
Plant Shutdown
84