IR 05000440/2024002
| ML24214A188 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/08/2024 |
| From: | Hartman T NRC/RGN-III/DORS/RPB2 |
| To: | Penfield R Vistra Operations Company |
| References | |
| IR 2024002 | |
| Download: ML24214A188 (1) | |
Text
SUBJECT:
PERRY NUCLEAR POWER PLANT - INTEGRATED INSPECTION REPORT 05000440/2024002
Dear Rod Penfield:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Perry Nuclear Power Plant. On July 11, 2024, 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 did not involve a violation of NRC requirements.
A licensee-identified violation which was determined to be of very low safety significance and Severity Level IV 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 violation or the significance or severity of the violation 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 Perry Nuclear Power Plant.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 Perry Nuclear Power Plant.
August 8, 2024 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, Thomas C. Hartman, Acting Branch Chief Reactor Projects Branch 2 Division of Operating Reactor Safety Docket No. 05000440 License No. NPF-58
Enclosure:
As stated
Inspection Report
Docket Number:
05000440
License Number:
Report Number:
Enterprise Identifier:
I-2024-002-0063
Licensee:
Vistra Operations Company, LLC
Facility:
Perry Nuclear Power Plant
Location:
Perry, OH
Inspection Dates:
April 01, 2024, to June 30, 2024
Inspectors:
J. Beavers, Senior Resident Inspector
V. Myers, Senior Health Physicist
T. Ospino, Resident Inspector
Approved By:
Thomas C. Hartman, Acting Branch Chief
Reactor Projects Branch 2
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Perry Nuclear Power 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
Shutdown Due to Exceeding Technical Specification Unidentified Reactor Coolant System Leakage Limit Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green FIN 05000440/2024002-01 Open/Closed
[P.3] -
Resolution 71153 The inspectors identified a finding of very low safety significance (Green) for the licensee's failure to properly identify and scope work on the A reactor recirculating pump suction valve packing located in the drywell into refueling outage 19.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000440/2024-001-00 LER 2024-001-00 for Perry Nuclear Power Plant,
Operation of the Residual Heat Removal Loops B and C Alternate Keep Fill Configuration Was Prohibited by Technical Specifications and Resulted in an Unanalyzed Condition 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On May 11, 2024, the plant shutdown for a planned maintenance outage to replace B reactor recirculating pump seals and restored to full power operation on May 18, 2024. On May 23, 2024, the plant shut down for a forced maintenance outage to address identified and unidentified reactor coolant system leakage in the drywell and was restored to full power on May 27, 2024. The plant operated at or near rated thermal power for the remainder of the inspection period with occasional power derates related to heat and humidity.
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
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (2 Samples)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from total solar eclipse on April 8, 2024.
- (2) The inspectors evaluated the adequacy of the overall preparations for summer readiness to protect risk-significant systems from heat related issues on week of June 17, 2024.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Review of the residual heat removal (RHR) B alignment at the 574 elevation on April 24, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 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 1AB-1E, RHR B 574' elevation room on April 24, 2024 (2)fire zone 1RB-1C-1C, drywell at the 599' and 583' elevations on May 12, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the quarterly announced fire drill on April 16, 2024.
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 plant shutdown on May 10 and 11, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated an evaluated simulator scenario on April 9, 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)steam bypass control system issues following maintenance outage on May 12, 2024 (2)reactor water cleanup delta flow timer on May 11, 2024
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
- (1) B reactor recirculating pump seal replacement on May 11 through 17, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 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)remote shutdown operability test, work order #200902588
- (2) B emergency service water pump outage, work order #20090138 (3)shutdown risk assessment for maintenance outage beginning May 11, 2024 (4)shutdown risk assessment for maintenance outage beginning May 23, 2024 (5)division 3 emergency diesel generator activities due to the water intrusion during the storm on May 18, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
(1)bus XH11 undervoltage relay
- (2) E intermediate range monitor operability assessment (3)steam bypass pressure control system functional assessment (4)reactor water cleanup system delta flow isolation functional assessment (5)high pressure core spray operability assessment (6)division 3 emergency diesel generator undervoltage relay operability assessment (7)division 3 emergency diesel generator water intrusion operability assessment
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1)permanent modification implementation of procedural allowance for use of alternate keep fill regarding residual heat removal system on April 11, 2024 (2)review of engineering change package 24-1051-001 on May 14, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated the planned maintenance outage activities associated with B reactor recirculating pump seal leakage from May 11 to May 19, 2024.
- (2) The inspectors evaluated the forced maintenance outage activities associated with A reactor recirculating pump suction valve packing leak from May 23 to May 27, 2024.
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) (7 Samples)
(1)division 1 T9 isolation transformer replacement on January 26, 2024
- (2) G average power range monitor 24-volt power supply replacement of on May 8, 2024
- (3) A reactor feed pump turbine solenoid valve replacement between May 12 and 17, 2024
- (4) B control rod drive pump maintenance finished on May 14, 2024
- (5) B reactor recirculating pump seal replacement finished on May 16, 2024
- (6) A reactor recirculating pump suction valve leak during forced cold shutdown between May 23 and May 26, 2024 (7)division 1 emergency diesel generator lube oil temperature module replacement on June 11, 2024
Surveillance Testing (IP Section 03.01) (6 Samples)
(1)remote shutdown operability test (24 months) between April 16 and 18, 2024 (2)breaker EH1205 ground over current calibration on April 24, 2024 (3)transformer LH-1-B deluge testing on May 12, 2024 (4)reactor cooldown surveillance on May 11, 2024 (5)reactor core isolation cooling valve and flow controller position verification surveillance on May 15, 2024 (6)transformer LH-1-C deluge testing on May 13, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) D and E main steam isolation valve closure channels response time surveillance (24 month) on April 10, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
(1)containment atmospheric monitoring isolation valves testing on May 12, 2024
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (2 Samples)
The inspectors evaluated:
(1)multifacility tabletop drill on April 30, 2024
- (2) multifacility tabletop drill on June 11,
RADIATION SAFETY
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
(1)control room emergency recirculation ventilation system
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) auxiliary building sump ventilation
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (2 Samples)
The inspectors evaluated the following internal dose assessments:
- (1) internal dose assessment for CR-2023-01912 (2)internal dose assessment for CR-2023-01488
Special Dosimetric Situations (IP Section 03.04) (2 Samples)
The inspectors evaluated the following special dosimetric situations:
- (1) use of effective dose equivalent for reactor water clean up work (2)declared pregnant workers
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) Unit 1 (April 1, 2023, through March 31, 2024)
MS06: Emergency AC Power Systems (IP Section 02.05) (1 Sample)
- (1) Unit 1 (April 1, 2023, through March 31, 2024)
MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) Unit 1 (April 1, 2023, through March 31, 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) B recirculating pump seal (2)nuclear closed cooling system effluent pathway (3)tritium presence in groundwater (4)steam head stud assembly modification pins
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program to identify trends in work planning and execution issues that might be indicative of a more significant safety issue.
- (2) The inspectors reviewed the licensees corrective action program to identify trends in engineering procedural compliance 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 loss of reactor water cleanup safety function and licensees response on May 16, 2024.
- (2) The inspectors evaluated the required shutdown due to exceeding technical specification unidentified reactor coolant system leakage limit and licensees response on May 23, 2024.
- (3) The inspectors evaluated the potential water intrusion into the division 3 emergency diesel generator during severe weather and licensees response on June 18, 2024.
Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.
- (1) LER 05000440/2024-001-00, (ADAMS Accession No. ML24172A162). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71111.18. This LER is closed.
- (2) LER 05000440/2024-002-00, (ADAMS Accession No. ML24179A014). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is closed.
INSPECTION RESULTS
Licensee-Identified Non-Cited Violation 71111.18 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.
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.
Violation: Technical Specification (TS) 3.5.1, ECCS-Operating, Condition C states, Two ECCS injection subsystems inoperable, Required Action C.1 states, Restore once ECCS injection/spray subsystem to OPERABLE status, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; Condition D states, Required Action and associated Completion of Condition A, B, or C not met, Required Action D.1 states, Be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, Required Action D.2 states, Be in MODE 4, in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, the plant was not in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the completion time for TS Limiting Condition for Operation (LCO) 3.5.1-required action C.1 was not met on March 30th, 2024, at 8:15 p.m. With the Division II waterleg pump, and by extension Division II B/C Residual Heat Removal (RHR) operability not being restored until March 30th, 2024, at 8:32 p.m. that period was greater than the allowed completion time by the limiting condition for operation provided in TS LCO 3.5.1.
Violation: 10 CFR 50.59(d)(1) requires the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to 10 CFR 50.59(c). These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section.
Per the Updated Safety Analysis Report, Section 6.3.2.2.5, use of the condensate storage tank water head, that is alternate keepfill, is only allowed for initial fill of the residual heat removal system and does not mention its use for maintaining the fill of the system. The waterleg pump is used to ensure that the discharge lines are full up to the isolation valves and to ensure no gas voids occur within the piping. The section also mandates that a high point vent procedure be performed after stopping and re-starting the waterleg pump.
Contrary to the above, between February 22, 2024, and April 11, 2024, the licensee failed to maintain records of changes in procedures made pursuant 10 CFR 50.59(c). Specifically, the licensee failed to provide a written evaluation to provide a basis for why Revision 80 of SOI-E12, Residual Heat Removal System, which removed the requirement for RHR to be declared inoperable when on alternate keepfill, did not require prior NRC approval. The licensee did not provide a basis for why the change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
Significance: Green. The inspectors screened the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 - Mitigating Systems Screening Questions, section A, Mitigating SSCs and PRA Functionality (except Reactivity Control Systems). The inspectors answered yes to question 1 as probability risk assessment functionality was maintained and that resulted in screening the issue to Green.
Severity: Severity Level IV. Since the violation resulted in a condition that was determined to have very low safety significance (Green), the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d.2 of the NRC Enforcement Policy.
Corrective Action References: CR-2024-03160 Observation: B Recirculating Pump Seals 71152A The inspectors performed an in-depth review of the B reactor recirculating pump seal and impact on drywell identified and unidentified leakage rates, the associated apparent cause evaluation, and implemented and planned corrective actions. The B reactor recirculating pump seal pressures and drywell leakage rates degraded from December 2023 to May 2024.
The licensee developed an Operational Decision-Making Instruction (ODMI) procedure to ensure trending and termination criteria were established. In May 2024, a planned maintenance outage was conducted starting May 11, 2024, and the seal was successfully replaced. The plant was restored to operation on May 17, 2024. Failure analysis included seal post-mortem disassembly, inspections, pictures, vendor discussions, comparisons with industry seal documentation and an apparent cause evaluation in accordance with the licensees corrective action program.
The licensees cause evaluation concluded that the B reactor recirculating pump seal damage resulted from particulate in the seal purge water passing through the filters into the seal cavities and scratching the highly polished seal faces. Once scratched, water leaked through from the high-pressure side to the low-pressure side, started to flash to steam and cut the softer carbon seal face. Steam cutting resulted in more leakage, more leakage results in more damage in a self-perpetuating, degrading trend. An Equipment Failure Analysis Checklist (EFAC), pointed to an inadequate design of seal purge filters. Current vendor recommendations best practices documents call out filters to remove particulate sized 1 micron and larger. Existing filters were 3-5 micron absolute. Multiple other issues such as pump shuttling, stray currents/residual magnetism, and low-pressure operation were identified as contributors to poor mechanical seal health.
The corrective actions developed include an engineering modification to change seal purge filters from existing 3-5 micron absolute to a filter 1 micron or better to meet recently developed industry best practices and vendor recommendations. An engineering evaluation request, EER #601451631, was submitted to engineering on May 21, 2024, regarding this plant modification. Additionally, a procedure change request was submitted to minimize reactor recirculation pump operating time with the reactor vessel at low pressure
(<50 psig).
The inspectors determined that the issues associated with the reactor recirculating pump seals were not reasonably within the licensees ability to foresee and correct and not a performance deficiency. The licensees effort in the apparent cause evaluation, including the corrective actions, were determined to be adequate, and the inspectors identified no more-than minor findings or violations associated with the product.
Observation: Shroud Head Stud Assembly Modification Pins 71152A The inspectors performed an in-depth review of the shroud head stud assembly modification (SHSAM) anti-rotating/locking pins, the associated Operational Decision Making Instruction (ODMI) procedure, and the impact of both the May 11, 2024, and May 23, 2024, reactor plant cold shutdown maintenance outages on the pins. These modification pins replaced the stud locking function originally provided by the shroud head stud bolts.
During refueling outage 19 in 2023, two of the 16 active SHSAM pins showed significant wear. An evaluation was performed by General Electric Hitachi (GEH) engineering that concluded the design condition for the SHSAM connections will be met as long as reactor water temperature does not reduce below 150 degrees Fahrenheit (F). Given the potential necessity to reduce reactor water temperature below 150 degrees F in support of a forced or planned maintenance outage prior to the next refueling outage in 2025, the GEH evaluation indicated that there could be preload assurance challenges to the SHSAM studs at locations
- 10 and #11.
The licensee-developed ODMI procedure concluded that if there was an unplanned outage requiring reactor water temperature to be reduced below 150 degrees F, the licensee management would have to make a decision as to whether or not to perform inspections of the affected SHSAM bolts to check for any loss of preload. Both May 2024 reactor plant cold shutdown maintenance outages met this threshold. Option one was to remove the reactor head and inspect the SHSAM pins in question. The second option was to implement monitoring criteria to identify degrading steam separator performance or equipment abnormalities during reactor plant startup. Both outage recoveries used the second option.
The licensee decisions were noted to be not without risk, as no analysis exists justifying less than 16 active SHSAMs precluding separator movement during a seismic event for example.
The inspectors reviewed the ODMI, GEH Engineering Evaluation 007N6957, and the associated licensee technical assessments for the SHSAM pin issue. The inspectors also observed both reactor startup evolutions and noted no indication of jet pump abnormalities or recirculating pump suction temperature abnormalities during startup that would indicate shroud head or steam separator functional challenges and potential steam bypass. Though not a safety function, the failure of the SHSAM pins could lead to an initiating event sequence requiring immediate plant shutdown. The event sequence would have been sufficiently guided by existing off normal instruction procedures. The inspectors evaluated the issues associated with the steam head stud assembly modification pins due to the cold shutdown and identified no more-than minor findings or violations.
Observation: Trends in Work Planning and Execution 71152S The inspectors performed a semiannual review of a potential adverse trend in the licensees corrective action program to work planning and execution issues that might be indicative of a more significant safety issue.
Included in this sample were the following corrective action program documents:
- CR 2024-02730, NEIL Notification for Fire Impairment expected to exceed 90 days
- CR 2024-02868, Non-Critical PM Deep-In-Grace Indicator Projecting Negative Trend
- CR 2024-03125, Seismic Monitoring Inoperable Greater than 30 Days
- CR 2024-03280, Potential non-like-for-like oil from vendor in starting air compressor
- CR 2024-04401, NEIL Notification made for three Fire Impairments expected to exceed 90 days
- Multiple corrective action program entry items associated with technical support center uninterruptable power supply
- Multiple corrective action program entry items associated with emergency core cooling system water leg pumps During the inspection period, the inspectors evaluated more than a dozen instances of both safety and support system maintenance delayed or scheduled deep into grace periods. The tracking and prioritization of such maintenance backlog challenges the resolution of issues and presents potential for the issues to be incorrectly prioritized or inadvertently forgotten altogether. Known maintenance items have challenged a number of the reactor oversight process safety cornerstones during the inspection period, and it is the inspectors perspective that the planning and execution of maintenance activities continue to add to an overall maintenance backlog that have a potential to postpone support and safety system impacting maintenance. The safety culture common language associated behavior was WP.1-Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the work process includes the identification and management of risk commensurate to the work. The inspectors completed the objectives of the inspection procedure and trended the behavior to consider potential safety culture weaknesses and discussed with the licensee about taking appropriate actions before significant performance degradation occurs as stated in NUREG-2165, Safety Culture Common Language. The issues addressed in this inspection were not identified to be more than minor since the licensee implemented appropriate contingencies and all equipment remained functional.
Observation: Trends in Engineering Procedural Compliance 71152S The inspectors performed a semiannual review of a potential adverse trend in the licensees corrective action program in engineering procedural compliance issues that might be indicative of a more significant safety issue.
Included in this sample were the following corrective action program documents:
- CR 2024-00602, Code Required Actions Not Taken For ESW Pump B Vibrations in Alert Range
- CR 2024-00605, Code Required Actions Not Taken For HPCS WATER LEG Pump Vibrations in Alert Range
- CR 2024-00794, Increased Vibrations on Emergency Closed Cooling Pump B
- CR 2024-00814, Pressure indicator 1E22-R0001 for High Pressure Core Spray suction found out of calibration
- CR 2024-01312, MMD-Drift Limit Exceeded for SIV-B21-T0138B
- CR 2024-01387, Code Required Actions Not Taken For LPCS & RHR A Waterleg Pump Vibrations in Alert Range
- CR 2024-01388, Code Required Actions Not Taken For FPCC Pump B Vibrations in Alert Range
- CR 2024-02321, Nonconservative Design Inputs used in Design Calculation B13-30
- CR 2024-02449, Code Required Actions Not Taken in a Timely Manner During the inspection period, the inspectors evaluated nine instances of both safety and support system maintenance code challenges. The inspectors followed the documented corrective action program issues from the first quarter of 2024 with observations of engineering training, both initial and continuing training, as well as engineering operational experience meetings. Engineering ownership of plant design in maintenance activities and in plant change evaluations was notably emphasized. The inspectors noted several instances where the necessary challenges regarding design basis were demonstrated by various engineering staff in both the maintenance and plant change processes. It is the inspectors perspective that the potential adverse trend in behavior that initiated this sample has largely been and continues to be addressed by the licensee and that better collaboration between the maintenance, operations, engineering, work planning, and management would accelerate the desired performance by all groups. The safety culture common language associated behavior was PA.3 Teamwork: Individuals and workgroups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, individuals demonstrate a strong sense of collaboration and cooperation in connection with projects and operational activities, and individuals work as a team to provide peer-checks, verify certifications and training, ensure detailed safety practices, actively peer coach new personnel, and share tools and publications. The inspectors completed the objectives of the inspection procedure and trended the behavior to consider potential safety culture weaknesses and discussed with the licensee about taking appropriate actions before significant performance degradation occurs as stated in NUREG-2165, Safety Culture Common Language. The issues addressed in this inspection were not identified to be more than minor.
Shutdown Due to Exceeding Technical Specification Unidentified Reactor Coolant System Leakage Limit Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green FIN 05000440/2024002-01 Open/Closed
[P.3] -
Resolution 71153 The inspectors identified a finding of very low safety significance (Green) for the licensees failure to properly identify and scope work on the A reactor recirculating pump suction valve packing located in the drywell into refueling outage 19.
Description:
On May 23, 2024, Perry Nuclear Power Plant, Unit 1 initiated a manual shutdown and entered a forced maintenance outage of the reactor plant when drywell unidentified leakage exceeded the Technical Specification (TS) limit. The event caused no potential degradation of the plants level of safety and no release of radioactive material. The resident inspector staff responded and was present, informed, and monitored the plant and personnel performance during the event. The NRC evaluated the event and documented its initial assessment in an NRC Management Directive -(MD) 8.3 document, -Decision Documentation for Reactive Inspection, available via ADAMS Accession Number ML24185A226. The NRC concluded a reactive inspection on this issue was not warranted since operator performance was as expected, and the estimated leakage remained within the makeup capability of permanently installed plant equipment.
As documented in Event Notification57136 submitted to the NRC on May 23, 2024, TS Action 3.4.5. condition B, unidentified Reactor Coolant System (RCS) leakage exceeds 5 gallons per minute, was entered on May 23, 2024, at 0000 with a required action to reduce leakage to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, due by 0400 on May 23, 2024. This required action to reduce leakage was not completed within the required time; therefore, a technical specification required shutdown was initiated and reported as a 4-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i).
After reactor shutdown and discovery of the source of the leak, the licensee estimated 12 gallon per minute (gpm) leakage from the A reactor recirculating pump suction valve stem packing and into the valves leak off detection line. This valve is designed for isolation during maintenance activities and has no protective features. From the leak off line, approximately 7.7 gpm leaked into the drywell general spaces through a flange leak in an inline, glass flow gauge and eventually condensed into the drywell floor drain sump as unidentified leakage.
The remaining leakage passed through the glass flow gauge via the normal leak off flow path and into the drywell equipment drain sump as identified leakage.
The plant was then placed in Mode 3, which is a hot shutdown condition, as required by technical specification actions. The licensee opened the reactor recirculating pump suction valve against its backseat and closed a manual isolation valve in the leak off line, upstream of the degraded flange, and inside the approved code boundary for normal operating reactor pressure and temperature. There was no indication that this degradation has any generic implication nor repetitive failure. From the initial leak identification and plant shutdown to the leak isolation and plant restoration, operators and equipment responded as expected. A post-maintenance inspection at normal operating pressure was completed with no indications of identified or unidentified leakage from the A reactor recirculating pump suction valve or supporting systems. A complete repair of the reactor recirculating pump suction valve packing is scheduled for the next refueling outage in 2025.
The inspectors determined appropriate TS conditions and actions were complied with, including entry into a 12-hour shutdown statement, as well as the regulatory notification requirements. The inspectors review of the licensees emergency plan identified no emergency action level thresholds were exceeded. No more-than minor findings were identified with the licensees performance and event response efforts.
The inspectors evaluated historical causal and corrective actions taken on the A reactor recirculating pump suction valve stem packing. In 2013, the initial indications of valve first stage packing issues were identified, with rising temperatures in the leak off line temperature monitor accompanied with very low volumetric changes in the identified leakage well below TS action levels. In 2020, the leak off line temperature again showed a rise in the first stage packing, with similar very low volumetric changes in the identified leakage well below TS action levels. During the refueling outage of spring, 2023, the valve in question remained out of the priority work regarding repairs. The licensees outage scope identification and control procedure requires items that need to be done during an outage, such as tasks designed to mitigate plant transients and items that cannot be repaired during operation to be included in the outage. Given the history and potential of this valve, the inspectors identified that NOBP-OM-4009, Outage Scope Identification and Control, was not appropriately applied to prevent this TS forced shutdown event. The licensees root cause investigation came to similar conclusions and the inspectors identified no issues with the licensee's root cause.
Corrective Actions: The licensee has planned a number of corrective actions, which include:
- develop and implement an engineering change package to eliminate leak off line for 1B33F0023A in conjunction with eliminating 3-stage packing configuration in refueling outage 1R20, and
- develop and implement a site valve packing preventive maintenance program for inaccessible locations while online.
Corrective Action References: CR-2024-04658
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to repair the degrading A reactor recirculating pump suction valve packing before causing excessive unidentified reactor coolant system leakage into the drywell and requiring a technical specification shutdown of the reactor plant was 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 inspectors compared the issue to the examples and guidance in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, the failure to address the degrading packing caused a technical specification required reactor plant shutdown and is similar to example 4.g of 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. The inspectors screened the finding using Exhibit 3 - Barrier integrity Screen Questions, section B, Reactor Coolant system (RCS) Boundary. The inspectors answered no to the question, which results in screening the issue to Green.
Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, deferrals of corrective actions are minimized; when required, due dates are extended using an established process that appropriately considers safety significance.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Observation: Water Intrusion into the Division 3 Emergency Diesel Generator During Severe Weather 71153 On June 18, 2024, Perry Nuclear Power Plant, Unit 1, entered the off normal instruction for severe weather. During a down pour of rain, water from the diesel generator building roof and exhaust hallway was able to make its way down the exhaust manifold into the division 3 diesel generator room. As part of the off normal instructions, plant equipment was assessed, and water was observed running down the division 3 diesel engine and on the electrical portion of the generator. The impacted diesel generator was in standby readiness at the time.
The event caused no other potential degradation the plants level of safety, and no release of radioactive material to the environment. The resident inspector staff responded and was present, informed, and monitored the plant and personnel performance during event decisions and actions. The NRC evaluated the event and documented its initial assessment in MD 8.3 Decision Documentation for Reactive Inspection available via ADAMS Accession Number ML24199A151. The NRC concluded a reactive inspection for this issue was not warranted since the resident inspector had already obtained a significant amount of information regarding the event and its potential causes, the event impact was limited to a single train of emergency power and the system it supplied, and operator performance during and following the event was expected.
As documented in Event Notification 57181 submitted June 19, 2024, Technical Specification (TS) Action 3.8.1 condition B, one required diesel generator inoperable, was entered on June 18, 2024, at 1640, with a required action to restore to operable status, due by 1640 on June 21, 2024. Therefore, this condition was reported as an 8-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). However, the notification was reported about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after the event rather than the required 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The inspectors and regional NRC staff maintained observation and understanding of the events during the entire sequence and were not impacted in their regulatory oversight functions.
While in the TS required actions, the division 3 diesel generator was removed from service, tagged out, visually inspected, and electrical tested. The generator was verified dry and electrical insulation was determined to be acceptable and capable of providing the proper insulation to ground to allow the diesel generator to run without compromising cables or capabilities to perform its safety function. On June 20, 2024, the division 3 diesel generator was declared operable and returned to standby readiness.
The inspectors determined appropriate TS conditions and actions were complied with regarding event response, except for a delayed regulatory notification, which had no impact on regulatory response or decision making as the resident staff was present and regional management continuously appraised. Regarding the delayed notification, the station generated a corrective action item entry with actions to clarify in-house guidance regarding the notification regulatory requirements associated with these instrument functions and applicability. Given the significance of and circumstances surrounding the matter, the inspectors characterized this performance deficiency as a minor violation of 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The inspectors identified no more-than minor findings with the licensees event response and performance efforts; however additional NRC inspection regarding event cause and corrective actions is ongoing and will be documented in a subsequent report.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 11, 2024, the inspectors presented the integrated inspection results to R. Penfield, and other members of the licensee staff.
- On June 11, 2024, the inspectors presented the radiation protection inspection results to R. Penfield, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
24 NOP-WM-
2002
Certification of Summer Readiness
06/27/2024
Procedures
VLI-E12
Residual Heat Removal System
04/24/2024
Drywell
05/12/2024
Fire Plans
FZ 1AB-1E
Unit 1 - RHR - B System 574 - 10 Elevation
04/24/2024
Procedures
FPI-A-B02
Fire Brigade Drills
04/17/2024
CR 2024-04251
RWCU Delta Flow Timer Is Locked in Alarm Due to Incorrect
RWCU Feed Return Reading
05/11/2024
CR 2024-04281
C85/N32 Module #2 Tripped Red Light is on in The Control
Room Indicating Potential Problem with Turbine Control or
Steam Bypass System
05/17/2024
CR 2024-04389
Annunciator System Ground
05/15/2024
CR 2024-04417
Broke Condenser Vacuum Leading to 3.5 Hour Startup Delay 05/16/2024
CR 2024-04419
Procedural Issues Delayed Plant Mode Change
05/16/2024
Corrective Action
Documents
CR 2024-04534
Reactor Pressure Regulator Setpoint Lower Than Normal
05/19/2024
CR 2024-052358
Water Leakage in Division 3 Diesel Generator Room
06/18/2024
Corrective Action
Documents
CR 2024-05266
Late NRC Notification of a Condition Could Prevent
the Fulfillment of a Safety Function
06/19/2024
Procedures
NOP-OP-1005-01
Shutdown Defense in Depth for B Recirculation Pump Seal
Outage
Remote Shutdown Operability Test
04/16/2024
Work Orders
ESW B Start Time Faster Than LAIZ/CR
04/24/2024
CR 2024-03640
Under/Overvoltage Relay Trip Target Locked In
04/24/2024
CR 2024-04246
Intermediate Range Monitor E Indication Erratic and Not
Able to range properly
05/11/2024
CR 2024-04251
RWCU Delta Flow Timer is locked in Alarm due to Incorrect
RWCU Feed Return Reading
05/11/2024
CR 2024-04283
Steam Bypass Pressure Control C85 Two Pumps Running
Alarm Not Functional
05/12/2024
CR 2024-04391
HPCS Alternate Keepfill Basis/Regulatory Applicability
Determination Concerns
05/15/2024
Corrective Action
Documents
CR 2024-04405
RWCU Delta Flow Are High
05/15/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR 2024-04460
Reactor Water Cleanup Leak Detection was Bypassed Due
to Erratic Indications
05/16/2024
CR 2024-04460
Reactor Water Cleanup Leak Detection was Bypassed Due
to Erratic Indications
05/17/2024
CR 2024-04484
Erratic High RWCU Delta Flow Readings
05/17/2024
CR 2024-04950
Div 3 EDG UV Relay Out of Tech Spec Allowable Limits
06/05/2024
ICI-C-C51-0003
Intermediate Range Monitoring (IRM) Voltage Preamplifier
Calibration
Procedures
IMI-E06-0009
Filling and Venting RWCU Differential High Flow Instruments
05/16/2024
CR 2024-02947
Potentially Inadequate Review of RHR Alternate Keep-fill
Allowance in SOI-E12, Rev. 80
04/04/2024
Corrective Action
Documents
CR 2024-03160
CFR 50.59 Review Committee - Failed Product
04/11/2024
Engineering
Changes
ECP 24-1051-001 Removal of Saf-T-Climb Ladder in ESWPH
Procedures
CFR 50.59
Screen 24-00297
Residual Heat Removal System
CR 2020-09332
New Replacement T8 & T9 Transformer Brackets Not Same
as Existing
2/09/2020
CR 2022-08793
Unable to Locate Parts for Work Orders
(200471341 & 200471343)
11/16/2022
CR 2024-00642
Trip of Reactor Feed Pump Turbine B
01/24/2024
CR 2024-01619
CRD Pump B Oil Leak
2/29/2024
CR 2024-04181
False Division 1 DG Lube Oil Temperature High Alarm
05/09/2024
Corrective Action
Documents
CR 2024-04656
Post Walkdown Leak Identified in the Drywell
05/23/2024
Engineering
Changes
ECP 21-0186-012 Change The GR-5 (50G) Set Point Time Delay From
Cycles to 6 Cycles for Relay PY-1R22Q0714
Engineering
Evaluations
EER 601451897
Provide Backseat Torque-Recirc Suction
05/24/2024
GEI-0001
Performing Insulation Resistance Checks
GEI-0001
Performing Insulation Resistance Checks
01/25/2024
GEI-0007A
Instructions For Cable and Wire Terminations
01/25/2024
GEI-0104
Maintenance and Calibration of Ground Fault Relays
Type GR-5
Procedures
SVI-B21-T1176
RCS Heatup and Cooldown Surveillance
01/26/2016
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
SVI-C51-T0030G
APRM G Channel Calibration for 1C51-K605G
SVI-C61-T1201
Remote Shutdown Panel 1C61-P001 Control
Operability Test RHR A, ESW A, ECC A, and
Division 1 Diesel Generator
04/16/2024
SVI-E31-T0101-A
RCIC Steam Supply Pressure Low Channel A Calibration for
SVI-E51-T1269
RCIC System Valve and Flow Controller Position Verification
(IPO-36) Replace PS413
05/08/2024
PTI-P54P0064B LH-1-B Spray Flow Test (24M) Water Spray
Flow Test for Unit 1 Interbus Transformer B
05/12/2024
PTI-P54P0064C LH-1-C Spray Flow Test (24M) Water Spray
Flow Test for Unit 1 Interbus Transformer C
05/13/2024
SVI-E51T1269 RCIC Vlv & Flow Controller (31D) RCOIC
System Valve and Flow Controller Position Verification
04/15/2024
SVI-C71T0253D 1 (24M-STB) MSIV Closure Channel D and
E RPS Response Time For 1C71A-k3D and 1C71A-K3E
04/10/2024
SVI-E31T0101A 1 (24M) RCIC Steam Supply Pressure Low
Channel A Calibration For 1E31-N085A 4/18/2024
04/18/2024
No Work Description Detailed
04/25/2024
SVI -SVI-D23T2001 1 (92D) (C/S) Containment Atmosphere
Monitoring Isolation Valve Operability Test
05/12/2024
Replace RFPT A Solenoid Valve, 1N27F0406A,
1N27F0408A, 1N27F0412A and, 1N27F0413A
Cables in the Week Ending
05/19/2024
Determine Cause/Repair CRD Pump Oil Leak
05/14/2024
SUPERSEDED
03/14/2024
Determine the Cause of the False Div 1 DG LO Temp. HI
Alarm and Repair/Replace Parts as Required
06/12/2024
Work Orders
Repair Leak at Site Flow Indicator
05/24/2024
Corrective Action
Documents
CR-2023-09046
SCBA Failed Function Test
2/08/2023
Corrective Action
Documents
Resulting from
CR-2024-03966
Use Of Zero Value in Procedures Related to HEPA and
Vacuum Units
05/02/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Inspection
Posi3 USB Test Results
09/06/2023
G-GEN-
SCBA_FEN
M7 Firehawk SCBA Respirator Qualification Records for
Various Individuals
04/02/2024
Miscellaneous
PNPP 8269
Respirator Inspection Records
2/02/2024
NOP-OP-4303
Respirator Quantitative Fit Test
Procedures
NOP-OP-4310
FireHawk M7 Self Contained Breathing Apparatus
200864338
Control Room Emergency Recirculation Subsystem A Flow
and Filter Operability Test
2/14/2023
Work Orders
200864339
Control Room Emergency Recirculation Subsystem B Flow
and Filter Operability Test
2/28/2022
Corrective Action
Documents
CR-2024-01169
Higher Than Expected Shallow and Eye Dose On TLD
2/14/2024
Corrective Action
Documents
Resulting from
Inspection
CR-2024-02892
Potential Issues with NRC Form 5s When EDEx Is Involved
04/03/2024
NOP-OP-4204-04
Effective Dose Equivalent Dose Determination Form
2/15/2024
NOP-OP-4205-02
TLD/SRD Deviation Investigation Report
2/01/2024
Miscellaneous
RPS-23-006
EDEX Plan for RWCU Repair
NOP-OP-4204
Special External Exposure Monitoring
NOP-OP-4205
Dose Assessment
Procedures
NOP-OP-4206
Bioassay Program
230357
230505
Scaffolding Work in Drywell
Radiation Work
Permits (RWPs)
230903
IVVI Activities
Corrective Action
Documents
CR 2024-04257
Reactor Recirculation Pump B Seal Pressures and Drywell
Leakage Rates Have Degraded from December 2023 to
May 2024 Resulting a Plant Outage for Seal Replacement
05/11/2024
Engineering
Evaluations
601449458
De-tensioned SHSAM Risks
05/08/2024
CR 2024-04458
Late NRC Notification of Loss of Safety Function
05/16/2024
Corrective Action
Documents
CR 2024-04458
Late NRC Notification of Loss of Safety Function
05/16/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR 2024-04460
Reactor Water Cleanup Leak Detection was Bypassed Due
to Erratic Indications
05/14/2024
CR 2024-04484
Erratic High RWCU Delta Flow Readings
05/17/2024
Reactor Water Cleanup System Isolation Channel Inoperable
05/16/2024
Technical Specification Required Shutdown
05/23/2024
Inoperability of Division 3 Diesel Generator Supporting High
Pressure Core Spray
06/19/2024
Miscellaneous
ODMI PY-24-03
SHSAM Pins 10 and 11 Degraded
1