IR 05000298/2023002
| ML23216A074 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/08/2023 |
| From: | David Proulx NRC/RGN-IV/DORS |
| To: | Dia K Nebraska Public Power District (NPPD) |
| References | |
| EPID I-2023-002-0010 IR 2023002 | |
| Download: ML23216A074 (28) | |
Text
August 08, 2023
SUBJECT:
COOPER NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000298/2023002
Dear Khalil Dia:
On June 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Cooper Nuclear Station. On July 11, 2023, the NRC inspectors discussed the results of this inspection with John Dent, Executive Vice President and Chief Nuclear Officer, 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. One Severity Level IV violation without an associated finding is also documented in this report. 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Cooper Nuclear Station.
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, David L. Proulx, Acting Chief Reactor Projects Branch C Division of Operating Reactor Safety Docket No. 05000298 License No. DPR-46
Enclosure:
As stated
Inspection Report
Docket Number:
05000298
License Number:
Report Number:
Enterprise Identifier:
I-2023-002-0010
Licensee:
Nebraska Public Power District
Facility:
Cooper Nuclear Station
Location:
Brownville, NE
Inspection Dates:
April 1, 2023, to June 30, 2023
Inspectors:
G. Birkemeier, Resident Inspector
K. Chambliss, Senior Resident Inspector
T. DeBey, Acting Resident Inspector
R. Kopriva, Acting, Resident Inspector
A. Siwy, Acting Senior Resident Inspector
H. Strittmatter, Emergency Preparedness Inspector
R. Williams, Operations Engineer
Approved By:
David L. Proulx, Acting Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Cooper Nuclear Station, 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
Preventive Maintenance Schedule Failed to Prevent an Age-Related Solenoid Valve Failure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2023002-01 Open/Closed EA-23-020 None 71153 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-citied violation (NCV) of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to implement a preventive maintenance schedule developed to specify inspection or replacement of parts that have a specific lifetime. Specifically, the main turbine bypass valve preventive maintenance schedule failed to identify degradation or replace parts that have a specified lifetime on the main turbine bypass valves fast-open permissive solenoid valve, resulting in an age-related failure of the solenoid valve and a subsequent reactor scram.
Safety Relief Valve Lift Settings Outside of Technical Specifications Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000298/2023002-02 Open/Closed Not Applicable 71153 The inspectors reviewed a self-revealed, Severity Level IV, non-cited violation of Technical Specification 3.4.3, Safety/Relief Valves and Safety Valves, for the licensees discovery through as-found test results that two of the eight Target Rock safety relief valve pilot assemblies failed to lift within the technical specifications lift setpoint requirements.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000298/2023-001-00 Valve Test Failures Result in Condition Prohibited by Technical Specifications 71153 Closed LER 05000298/2022-001-00 Secondary Containment Differential Pressure Perturbation Exceeds Technical Specifications Limit 71153 Closed
Type Issue Number Title Report Section Status LER 05000298/2022-004-00 Manual Reactor Scram and Group I Isolation due to Main Turbine Bypass Valve Failing Open 71153 Closed LER 05000298/2022-004-01 LER 2022-004-001 for Cooper Nuclear Station,
Manual Reactor Scram and Group I Isolation due to Main Turbine Bypass Valve Failing Open 71153 Closed
PLANT STATUS
Cooper Nuclear Station began the inspection period at rated thermal power. On May 8, 2023, power was lowered to approximately 55 percent for an unplanned reduction in reactor power due a perceived inoperability of the core spray system. The plant returned to rated thermal power on May 9, 2023. On May 15, 2023, power was lowered to approximately 83 percent for a planned rod insertion and maintenance activities. The plant returned to rated thermal power on May 15, 2023. On May 18, 2023, power was lowered to approximately 63 percent for core power management. The plant returned to rated thermal power on May 19, 2023. On June 2, 2023, power was lowered to approximately 83 percent for core power management.
The plant returned to rated thermal power on June 3, 2023. The unit remained at 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)division 1 emergency diesel generator on April 18, 2023 (2)division 2 250 Vdc battery system on April 26, 2023
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (3 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)turbine generator building electrical and instrumentation shop, non-critical switchgear room, operating floor, 932-foot elevation, on April 20, 2023 (2)cable spreading room, 918-foot elevation, on April 26, 2023 (3)division 1 emergency diesel generator room, elevations 917-foot 6-inches and 903-foot 6-inches, on May 9, 2023
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 quarterly downpower for core management on May 20, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated simulator licensed operator training on April 14, 2023.
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)turbine generator function TGF-F01 evaluation to remain (a)(2) on May 12, 2023 (2)maintenance work practices resulting in configuration control issues on May 12, 2023
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (6 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)planned Yellow online risk during emergency diesel generator 2 maintenance window on April 18, 2023 (2)planned Yellow online risk during 4160 V, bus 1F, undervoltage relay testing on May 2, 2023 (3)planned Yellow online risk during division 1 residual heat removal maintenance window on May 3, 2023 (4)planned Yellow online risk during high pressure coolant injection maintenance window on May 17, 2023 (5)emergent Yellow online risk due to low out of specification 250 Vdc battery cell voltage on June 2, 2023 (6)planned Yellow online risk during standby gas treatment division 1 maintenance window on June 8, 2023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)calibration and erratic indication concerns for drywell high range radiation monitors on April 10, 2023 (2)installation of two control rod drive hydraulic accumulator solenoid valves without a design equivalent change on April 14, 2023 (3)multiple areas of corrosion identified on the inlet and outlet cover of diesel generator jacket water cooler exceeding minimum wall thickness on April 19, 2023.
(4)degradation of coal tar coating of service water piping to the emergency diesel generators on May 1, 2023 (5)residual heat removal motor operator valve disconnect switch fuse size discrepancy between installed fuse and design documents specification on May 5, 2023 (6)core spray division 1 logic circuit relay contacts failure to close on May 9, 2023 (7)reactor core isolation cooling steam supply valve calculation errors on June 29, 2023
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (3 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1)high pressure coolant injection auxiliary oil pump pressure switch modification on May 10, 2023 (2)residual heat removal pump C reservoir drain modification on May 11, 2023 (3)temporary configuration change installation of jumper across low out-of-specification battery cell in 250 Vdc division 1 subsystem on June 15, 2023
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) (6 Samples)
(1)control rod drive 06-31 instrument block shutoff testing following cartridge replacement on April 21, 2023 (2)standby gas treatment division 1 maintenance window post work testing on April 21, 2023 (3)emergency diesel generator division 2 maintenance window post maintenance test on May 4, 2023 (4)service water booster pump C functional test following maintenance on May 8, 2023 (5)residual heat removal division 1 maintenance window post maintenance test for motor operated valve on June 15, 2023 (6)individual battery cell replacement for inoperable 250 Vdc division1 battery cell on June 15, 2023
Surveillance Testing (IP Section 03.01) (5 Samples)
(1)containment high range area monitors (RM-40A/B) on April 10, 2023 (2)standby liquid control quarterly operability test on April 14, 2023 (3)augmented off-gas hydrogen monitor functional test on April 21, 2023 (4)emergency diesel generator division 1 monthly test on May 10, 2023 (5)residual heat removal division 1 flow test on June 1, 2023
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)residual heat removal division 2 quarterly in-service test on May 5, 2023
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
(1)diverse and flexible coping strategies (FLEX) equipment annual and quadrennial maintenance window on May 24, 2023
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
- (1) The inspectors evaluated the maintenance and testing of the alert and notification system between January 1, 2021, and May 12, 2023.
71114.03 - Emergency Response Organization Staffing and Augmentation System
Inspection Review (IP Section 02.01-02.02) (1 Sample)
- (1) The inspectors evaluated the readiness of the Emergency Preparedness Organization between January 1, 2021, and May 12, 2023. Inspectors also evaluated the licensee's ability to staff their emergency response facilities in accordance with emergency plan commitments.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated the 10 CFR 50.54(q) emergency plan change process and practices between November 1, 2020, and January 31, 2023. The evaluation reviewed screenings and evaluations documenting implementation of the process. In addition, the inspectors evaluated:
Emergency Plan, revision 80, effective July 28, 2022
Emergency Plan, revision 81, effective December 6, 2022
Emergency Plan Implementing Procedure 5.7.20 Protective Action Recommendations, Revision 33, effective April 6, 2022
Alert and Notification System Design Report, Revision 18, effective March 1, 2022 This evaluation does not constitute NRC approval.
71114.05 - Maintenance of Emergency Preparedness
Inspection Review (IP Section 02.01 - 02.11) (1 Sample)
- (1) The inspectors evaluated the maintenance of the emergency preparedness program between January 1, 2021, and May 12, 2023. The evaluation reviewed evidence of completing various emergency plan commitments, the conduct of drills and exercises, licensee audits and assessments, and the maintenance of equipment important to emergency preparedness.
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)
(1)emergency preparedness drill on April 13,
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) April 1, 2022, through March 31, 2023
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
- (1) April 1, 2022, through March 31, 2023
EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)
- (1) April 1, 2022, through March 31, 2023 EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)
- (1) April 1, 2022, through March 31, 2023 EP03: Alert And Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)
- (1) April 1, 2022, through March 31, 2023
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)radiological and flood barrier control process for review on May 5, 2023 (2)causal analysis review of fire event in the cable spreading room and review of the corrective actions on June 20, 2023
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000298/2022-001-00, Secondary Containment Differential Pressure Perturbation Exceeds Technical Specifications Limit (ML22201A521). The circumstances surrounding this LER and a minor violation are documented in the Inspection Results section of this report. This LER is closed.
- (2) LER 05000298/2023-001-00, Valve Test Failures Result in Condition Prohibited by Technical Specifications (ML23128A133). The circumstances surrounding this LER and a Severity Level IV, non-cited violation are documented in the Inspection Results section of this report. This LER is closed.
- (3) LER 05000298/2022-004-00, Manual Reactor Scram and Group I Isolation due to Main Turbine Bypass Valve Failing Open (ML23045A162) and LER 05000298/2022-004-01, Manual Reactor Scram and Group I Isolation due to Main Turbine Bypass Valve Failing Open (ML23132A132). The circumstances surrounding these LERs and a Green, non-cited violation are documented in the Inspection Results section of this report. These LERs are closed.
INSPECTION RESULTS
Preventive Maintenance Schedule Failed to Prevent an Age-Related Solenoid Valve Failure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000298/2023002-01 Open/Closed EA-23-020 None 71153 The inspectors reviewed a self-revealed finding of very low safety significance (Green) and an associated non-citied violation (NCV) of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to implement a preventive maintenance schedule developed to specify inspection or replacement of parts that have a specific lifetime. Specifically, the main turbine bypass valve preventive maintenance schedule failed to identify degradation or replace parts that have a specified lifetime on the main turbine bypass valves fast-open permissive solenoid valve, resulting in an age-related failure of the solenoid valve and a subsequent reactor scram.
Description:
On December 15, 2022, Cooper Nuclear Station lowered power to a band of 10 to 15 percent with bypass valves in pressure control. This was in response to elevated hydrogen levels on the main turbine deck due to a hydrogen leak on a main turbine bushing.
On December 16, 2022, bypass valve 1 failed open and bypass valves 2 and 3 closed to attempt to maintain reactor pressure. The control room responded by starting a second electro-hydraulic control fluid pump to restore the lowering fluid pressure and recover the bypass valves. The failed open bypass valve maintained a condition where pressure control was not possible for the current power level with the mode selector switch still in RUN.
While attempting the recovery, operators noticed that reactor pressure had fallen below the low-pressure scram action setpoint listed for the Digital Electro-Hydraulic (DEH) Fluid System Trouble alarm (900 psig). Operators inserted a manual scram with reactor pressure at approximately 860 psig. At nearly the same time the manual scram was inserted, a full Group 1 isolation (full closure of the main steam isolation valves) occurred automatically on main steam line low pressure (>=835 psig). The total length of time from the bypass valve failing open to the scram and full Group 1 isolation was 2 minutes, 36 seconds. The licensee was able to regain pressure control approximately 70 minutes after the full Group 1 isolation by manually opening the main steam line drains.
Troubleshooting identified that bypass valve 1 failed open due to a failed fast-open permissive solenoid valve (SOV). The function of the fast-open permissive SOV valve is to remain energized and closed when reactor power is below 115 Mwe with the output breaker to the step-up transformer to the main grid closed (or anytime when the output breaker is open).
The valve is controlled by the DEH control system. There is a single SOV associated with each bypass valve. When the SOV closes (normal position at power), the valve will keep the associated dump valve pressurized and allow the bypass valve to modulate its position based on the DEH demand to the bypass valves servo positioner. When the SOV is de-energized, the dump valve can be depressurized via the governor valve emergency trip (GVET) header sending supply oil to the bottom of the bypass valve causing the bypass valve to go full open.
Since the turbine was offline due to hydrogen leak repairs, the GVET header was depressurized causing the bypass valve to go full open upon the SOVs failure.
The licensee removed and replaced the SOVs for all three bypass valves. Each SOV was sent to a lab for further failure analysis. The lab analysis report did not result in a conclusive cause to the SOV failure.
The failed SOV was removed and showed signs of blistering and overheating on the solenoids coil. The SOV was installed in 2008 (along with the other two SOV valves for the other two bypass valves). Additionally, when the three valves were removed, engineering compared them to pictures of the valves that were removed in 2008. Engineering noticed the valves did not appear to be like-for-like. Engineering has reached out to the solenoid valve vendor for a detailed analysis of the differences between the valves.
Technical Specification (TS) 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978. NRC Regulatory Guide 1.33, revision 2, appendix A, section 9, provides recommendations for "Procedures for Performing Maintenance." Part b of section 9, states, in part, that "preventative maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime."
The inspectors reviewed applicable station procedures, work orders, corrective action documents, and preventive maintenance documents for performing maintenance and identified the following:
The preventive maintenance (PM) frequency for SOV replacement was established in 2010.
Initially, the recommendation from system engineering was every six refueling cycles (RE) to establish a RE 6/9-years PM frequency. In 2014, due to the licensee changing refueling outages from 18-months to 24-months, system engineering recommended a change to RE 5/10-years PM frequency. The sites preventive maintenance ownership group challenged this recommendation. System engineering then returned with a recommendation of RE8/16-years PM frequency.
Since there were no vendor recommendations available for a PM frequency, the system engineering team used EPRI reference document Valve - Solenoid Operated - SOV, revision 0, for guidance to determine the PM frequency. System engineering determined the most limiting component for the SOVs were elastomers. Previous operating experience had demonstrated that the elastomers had functioned for up to 18-years.
While evaluating the EPRI guidance, system engineering determined the SOV would be classified under Valve - Solenoid Operated - SOV, revision 0, as non-critical, low duty cycle, severe environment valve. For this classification, the EPRI document has no recommended replacement frequency.
-
The EPRI document defines Critical as functionally important, e.g., risk significant, required for power production, safety-related, or other regulatory requirements. The SOVs affect the operation of the turbine bypass valves which are controlled by Cooper Nuclear Station Technical Specification 3.7.7, The Main Turbine Bypass System. As such, the SOVs are critical due to the other regulatory requirements.
The EPRI document recommends that for a classification of critical, low duty cycle, severe environment, to replace the valves on a frequency of every 5 years. The EPRI document also states coil insulation breakdown may occur between 10-15 years.
The licensee reached out to the industry for comparable PM histories. A licensee with similar solenoid valves performing a similar function informed CNS they have a PM frequency of 9 years.
Engineering Desktop Guide 98-03-02, revision 6, provides guidance on classifying components in accordance with AP-913, Equipment Reliability Process Description, revision 5.
Critical Class 1 components are those components that directly support the proper operation or completion of at least one Important Function. The failure of a Critical Class 1 component could result in the undesirable consequences that are Critical Class 1 criteria on the CCD Matrix.
The CCD Matrix consequence table includes a reactor or turbine trip as one of the production criticality consequences resulting from a loss of function of the component.
Critical Class 1 components are assigned a maintenance basis for which preventive maintenance can be performed to maintain reliability of the component at expected levels or increase the reliability of the component to a new higher level.
Procedure 3-CNS-DC-324, Preventative Maintenance Program, revision 7, establishes and defines the requirements for the performance and documentation of the preventive maintenance (PM) program for CRIT1, CRIT2, and CRITN components at CNS. Per the PM program procedure, the objective of a PM program is to prevent in-service failures of CRIT1 and CRIT2 plant equipment and maintain equipment in a satisfactory condition for normal and/or emergency use.
Additionally, Procedure 3-CNS-DC-324, Section 8.4, provides direction on providing a justification for a frequency extension that deviates from industry standards or EPRI PM basis document templates. The PM justification for the SOVs was developed in January 2014.
System engineering cited maintenance history for the recommended PM frequency of 16 years despite replacing the SOVs after 14 years in 2008. Thus, the engineers determined that a replacement PM frequency of 16-years was adequate. However, the solenoid valves installed in 2008 were not like-for-like to the previously installed solenoid valves. As such, using previous maintenance history to justify a 16-year replacement cycle was inadequate.
The SOVs were originally planned to be replaced during the refueling outage in October 2022; however, this replacement PM was deleted from the outage scope and planned to be completed in the next refueling outage in September 2024.
Based upon the above information the inspectors determined the following:
Contrary to station procedures for the preventive maintenance program, the turbine bypass valves fast-open permissive solenoid valve, which is a critical component, had a preventive maintenance frequency that deviated from industry standards and the justification for the frequency failed to consider a different model solenoid valve being installed to prevent an age-related failure of the solenoid valve as required by the stations TS, Regulatory Guide 1.33, and the stations preventive maintenance procedures.
Corrective Actions: The licensee replaced all three associated fast-open permissive solenoid valves and the station changed the preventive maintenance frequency to a more conservative timeframe of 8-years. Additionally, the licensee is planning to replace all the fast-open permissive solenoid valves during 2024 refueling outage.
Corrective Action References: condition report CR-CNS-2022-06812
Performance Assessment:
Performance Deficiency: Regulatory Guide 1.33, Revision 2, Appendix A, section 9, provides recommendations for "Procedures for Performing Maintenance." Part b of section 9, states, in part, that "preventative maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime." The inspectors determined that the licensee's preventive maintenance strategy failed to detect degradation of solenoid valve components during preventive maintenance tasks or to have tasks that would replace solenoid components before an age-related failure of the fast-open permissive solenoid valve occurred and was therefore a performance deficiency.
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. Specifically, the fast-open permissive solenoid valve failure resulted in the main generator bypass valve 1 failing open and resulting the operators having to insert a manual scram.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. T Using the Initiating Events screening questions for transient initiators, the inspectors determined that a detailed risk evaluation would be needed because the finding caused a reactor scram and a loss of mitigating equipment. Using the conditional core damage probability methodology for event-based risk assessment, the analyst ran this evaluation as a loss of condenser heat sink because the event resulted in closure of the main steam isolation valves.
The analyst noted that the solenoid was normally deenergized and it failed after about 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> of energized operation, so the analyst considered the solenoid had 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> of energized life until failure. Because the solenoid is only energized below 15 percent power, this failure mode would lead to conditions where the plant would have a lower decay heat load than a normal full power condition. The analyst reviewed traces from past main steam isolation valve closure events and observed that similar events with a significant power history of full power operations yielded significantly different plant response to the event being analyzed. Because the turbine bypass valve failed open, plant pressure was initially reduced to approximately 700 psig. The lower decay heat load brought about by the failure mode resulted in a low reactor vessel re-pressurization rate. The lower rate would give operators plenty of time to recognize the need for and to establish dumping steam to the condenser.
The analyst re-quantified basic event PCS-XHE-XL-LOCHS, Power Conversion System Recovery during Loss of Condenser Heat Sink, using the SPAR-H human reliability analysis methodology by using all nominal performance shaping factors to obtain a human error probability of 1.1E-2. This model change represented the failure probability for operators taking action to vent the reactor vessel by bypassing around the main steam isolation valves and dumping steam through the failed open turbine bypass valve. Second, the lower decay heat inherent to this event reduced the amount of water needed for reactor vessel inventory makeup. To account for this change, the analyst lowered the probability of basic event RCI-TDP-FR-TRAIN to 2.8E-2, which allowed early crediting of the control rod injection pumps for inventory makeup. Third, the analyst eliminated the top 5 cutsets which contained failures of containment venting because the low decay heat load inherent to the failure would not degrade containment conditions to the point of needing to vent containment. Incorporation of these three model modifications yielded a conditional core damage probability of 2.9E-7, characterizing the finding with very low safety significance (Green).
The analyst applied the applicable large early release frequency factors from Table 6.2, Phase 2 Assessment Factors - Type A Findings at Power, from Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, to the dominant sequences to estimate the conditional large early release probability at 4.2E-8, which is also of very low safety significance. Also, the analyst determined that external events would not add appreciable significance due to the low probability of having an external event coincident with the analyzed event. Dominant core damage sequences included losses of condenser heat sink events coincident with anticipated transients without scram, which were mitigated by the standby liquid control system and reactor recirculation pump trip function. The analyst used the SPAR model, version 8.80, ran on SAPHIRE software, version 8.2.7, to estimate the significance.
Cross-Cutting Aspect: None. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
TS 5.4.1.a requires, in part, written procedures shall be established, implemented, and maintained as covered in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Section 9.b specifies, in part, that preventive maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime.
Procedure 3-CNS-DC-324, Preventative Maintenance Program, revision 7, implements Regulatory Guide 1.33 and provides the requirements for establishing and documenting the preventive maintenance program for CRIT1 components.
Contrary to the above, from January 2014 to June 6, 2023, preventive maintenance schedules were not developed to specify inspection or replacement of parts that have a specific lifetime. Specifically, in accordance with Cooper Nuclear Station Procedure 3-CNS-DC-324, the licensee failed to establish a preventative maintenance schedule that would result in the replacement of the fast-open permissive solenoid valves for the main generator bypass valves prior to the end of their specified timeline. As a result, on December 16, 2022, the solenoid valve for the main generator bypass valve1 suffered a failure and resulted in a manual reactor scram and a full Group 1 isolation.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Safety Relief Valve Lift Settings Outside of Technical Specifications Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000298/2023002-02 Open/Closed Not Applicable 71153 The inspectors reviewed a self-revealed, Severity Level IV, non-cited violation of Technical Specification 3.4.3, Safety/Relief Valves and Safety Valves, for the licensees discovery through as-found test results that two of the eight Target Rock safety relief valve pilot assemblies failed to lift within the technical specifications lift setpoint requirements.
Description:
Licensee Event Report 05000298/2023-001-00, Valve Test Failures Result in Condition Prohibited by Technical Specifications, (ML23128A133), was associated with two of the eight Target Rock safety relief valve (SRV) pilot assemblies as-found setpoints being outside of the +/-3 percent setpoint band required for their operability. This was discovered between March 7 and March 8, 2023, during as-found testing on all eight SRV pilot assemblies that were removed during the fall 2022 refueling outage. The licensee discovered that the two SRV pilot valves stuck due to corrosion bonding. The licensee determined that these two SRVs were inoperable for an indeterminate time period from October 31, 2020, when the unit entered mode 2 (beginning of operating cycle) to October 1, 2022, when the unit entered mode 4 (beginning of refueling outage). The inspectors reviewed the licensee event report and determined that the report adequately documented the summary of the event including the cause and potential safety consequences. The inspectors also reviewed other documents that indicate that this type of failure is a known industry issue associated with this type of valve.
Corrective Actions: The licensee replaced all eight of the SRV pilot valve assemblies with refurbished valves during the fall 2022 refueling outage. The currently installed valves were certified, tested, and as-left values were verified to be within +/-1 percent of their setpoints. The licensee is tracking industry initiatives to address the known corrosion bonding phenomenon and is working on a technical specification amendment to address this issue.
Corrective Action References: condition reports CR-CNS-2023-01034, CR-CNS-2023-01057, and CR-CNS-2023-01066
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Enforcement:
Enforcement Policy, section 2.2.4, states that violations with no associated performance deficiency will be dispositioned using traditional enforcement. Therefore, operating reactor violations with no associated performance deficiencies should be assigned a severity level and are, thus, not described as findings or assigned a color (e.g., Green).
Severity: Traditional Enforcement is used to disposition violations with no associated Reactor Oversight Process performance deficiency, per Enforcement Manual, Section 3.10, Reactor Violations with No Performance Deficiencies. The inspectors reviewed this issue in accordance with Inspection Manual Chapter (IMC) 612 and the Enforcement Manual. The inspectors reviewed Section 6.1.d.1 of the Enforcement Policy and determined this violation was Severity Level IV because it was a failure to comply with a TS action requirement for an LCO in TS Section 3.0.
Technical Specification 3.4.3, Safety/Relief Valves (SRVs) and Safety Valves (SVs),condition A, requires that with one or more required SRVs or SVs inoperable, that the unit 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 /> and 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, from October 31, 2020, to October 1, 2022, with two required SRVs inoperable, the licensee failed to place the unit in mode 3 and mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> respectively.
Enforcement A: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Minor Violation 71153 Minor Violation: During inspector review of LER 05000298/2022-001-00, a minor violation of Cooper Nuclear Station Technical Specification 5.4.1.a was noted. Technical specification 5.4.1.a requires, in part, written procedures shall be established, implemented, and maintained covering the following activities: a. the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978. Regulatory Guide 1.33, revision 2, appendix A, paragraph 9.b requires, in part, preventive maintenance schedules should be developed to specifyinspection or replacement of parts that have a specific lifetime such as wear rings.
Contrary to the above, from 2002 to May 23, 2022, the licensee did not have a preventive maintenance developed to specify inspection or replacement of parts that have a specific lifetime such as wear rings. Specifically, the licensee did not include soft goods such as O-rings in the preventive maintenance plan for diaphragms for the exhaust vortex damper actuators that maintain secondary containment. On May 23, 2022, this resulted in a failure of the diaphragm for damper actuator for exhaust fan A failing leading to secondary containment to exceed the technical specification surveillance requirement 3.6.4.1.1 limit of -0.25 inches wg for 2 minutes. Secondary containment was subsequently re-established to operable status with no operator actions required.
Screening: The inspectors determined the performance deficiency was minor. The inspectors determined the violation to be minor because, in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Additional Issue Screening Guidance, traditional enforcement does not apply, and the performance deficiency (PD) does not meet any of the More-than-Minor (MTM) criteria. The PD did not meet any of the MTM criteria, because, in this instance, secondary containment was able to be restored without any operator actions.
Enforcement:
The licensee has entered the issue into their corrective action program as condition report CR-CNS-2022-02184 to restore compliance. This failure to comply with procedural requirements did not affect the barrier integrity cornerstone objective. As a result, this issue is of low safety significance and constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On May 18, 2023, the inspectors presented the emergency preparedness program (IPs 71114.02, 71114.03, 71114.04, and 71151) inspection results to Khalil Dia, Site Vice President, and other members of the licensee staff.
On May 25, 2023, the inspectors presented the emergency preparedness program (IP 71114.05) inspection results to Khalil Dia, Site Vice President, and other members of the licensee staff.
On July 11, 2023, the inspectors presented the integrated inspection results to John Dent, Executive Vice President and Chief Nuclear Officer, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-CNS-
22-03104, 2022-04745, 2022-06185, 2023-00687, 2023-
01169
Drawings
DWG E507,
Sheet 23
EE-STR-HPCI (ALOP) Aux Lube Oil Pump
Procedures
2.2.24.2
250 VDC Electrical System (DIV 2)
Procedures
2.2A.DG_DIV1
Standby AC Power System (Diesel Generator) Component
Checklist (DIV 1)
Procedures
2.2A_250DC.DIV2
250 VDC Power Checklist (DIV 2)
Procedures
2.2B.DG.DIV1
Standby AC Power System (Diesel Generator) Instrument
Valve Checklist (DIV 1)
Procedures
6.EE.605
250V Battery Service Test
Corrective Action
Documents
CR-CNS-
21-00051, 2021-00100, 2021-02419, 2021-03112, 2022-
221, 2022-03577, 2022-03831, 2022-05153, 2022-05335,
22-05941, 2023-00550, 2023-01044, 2023-01855
Fire Plans
CNS-FP-229
Fire Protection Plan
Fire Plans
CNS-FP-236
Fire Protection Plan
Fire Plans
CNS-FP-251
Fire Protection Plan
Fire Plans
CNS-FP-252
Fire Protection Plan
Fire Plans
CNS-FP-253
Fire Protection Plan
Procedures
0-BARRIER
Barrier Control Process
Procedures
0-BARRIER-
CONTROL
Control Building
Procedures
0-BARRIER-MISC
Miscellaneous Buildings
Procedures
0-BARRIER-
TURBINE
Turbine Building
Procedures
0.23
CNS Fire Protection Plan
Corrective Action
Documents
CR-CNS-
23-00596
Procedures
10.9
Control Rod Scram Time Evaluation
Procedures
2.0.3
Conduct of Operations
106
Procedures
2.1.22
Recovering from a Group Isolation
Procedures
2.4MC-RF
Condensate and Feedwater Abnormal
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
2.4OG
Off-Gas Abnormal
Procedures
2.4RR
Reactor Recirculation Abnormal
Procedures
5.3DC125
Loss of 125 VDC
Procedures
RPV Control (1-3)
Procedures
Alternative Level/Pressure Control
Procedures
Primary Containment Control (1-3)
Procedures
Secondary Containment Control (1-3)
Procedures
EPIPEALHOT
Corrective Action
Documents
CR-CNS-
22-06812, 2022-06814, 2022-06897, 2023-01453, 2023-
01956, 2023-02322
Miscellaneous
Maintenance Rule Function TGF-F01 Performance Criteria
Basis
Procedures
0.31.1
Configuration Control During Maintenance Activities
Procedures
2.0.3
Conduct of Operations
106
Procedures
2.2.3
161
Procedures
3-CNS-DC-324
Preventative Maintenance Program
Procedures
3.47.31
Periodic Surveillance and Preventative Maintenance
Program
Corrective Action
Documents
CR-CNS-
23-01782, 2023-01819, 2023-01955, 2023-02616, 2023-
2629
Miscellaneous
Protected Equipment Tagout DGB-1-DG2 WEEK 2316
Miscellaneous
Protected Equipment Tagout RHRA-2-RHR-SUBSYS-A WK
2318
Miscellaneous
Protected Equipment Tagout EDC1-1-250VDCA INOP WK
22 -A
Miscellaneous
Protected Equipment Tagout SGTA-1-SGT DIV 1 WEEK
23
Miscellaneous
DGB-1-DG2
Tagout, Week 2316
Procedures
0-CNS-FAP-OM-
031
Emergent Issue Response, Risk Classification, and
Oversight Determination
Procedures
0-CNS-WM-104
On-Line Risk Assessment
Procedures
0-PROTECT-EQP
Protected Equipment Program
Procedures
2.0.2
Operation Logs and Reports
23
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
6.1EE.302
4160V Bus 1F Undervoltage Relay and Relay Timer
Functional Test (DIV 1)
Procedures
6.HPCI.204
HPCI-SOV-SSV64 and HPCI-SOV-SSV87 IST Closure Test
Calculations
NEDC 91-079
System Level Design Basis Review for Reactor Core
Isolation Cooling System - MOVs, CED/EE Number: EE 08-
2
Calculations
BWR Owners Group Report on the Operational Design
Basis of Selected Safety - Related Motor-Operated Valves
09/1986
Calculations
NEDC: 95-003
Design Calculations Sheet for RCIC-MOV-MO131
Corrective Action
Documents
CR-CNS-
2015-01411, 2020-05244, 2022-06878, 2023-00960, 2023-
01174, 2023-01953, 2023-01961, 2023-01965, 2023-01976,
23-01977, 2023-01998, 2023-02201, 2023-02252, 2023-
263, 2023-02274, 2023-02896, 2023-03100
Drawings
DWG 3059,
Sheet 1
D.C. Panel Schedules
Drawings
DWG 3750,
Sheet 6
Annunciator Loop Diagram ANN-MUX-00
Drawings
DWG 777-3
3-900# Globe Valve-R.S. Press Seal Cast Carb STL -
Stellite Trim - B.W. Ends SMB-00 (15 Ft-lbs.) Motor
Operator
N03
Drawings
DWG 791E265,
Sheet 1
Core Spray System Elementary Diagram
Drawings
DWG 791E265,
Sheet 2
Core Spray System Elementary Diagram
Drawings
DWG A-1928-X
RCIC MOV-MO14 Oil Piping Diagram
10/31/2018
Drawings
DWG B7122-145
Valve SSPV UO6
Drawings
DWG E501,
Sheet 31A
Cooper Nuclear Station Integrated Control Circuit Diagram
RCIC-MOV-131 Steam Supply to RCIC Turbine
N01
Drawings
DWG F-2041
Cooper Nuclear Station Flow Diagram Reactor building Main
Steam System
11/09/2022
Drawings
DWG L-4002
RCIC Throttle Valve Linkage arrangement to Permit Valve to
be Reset Automatically, Without Oil Pressure
2/10/1976
Drawings
DWG P-3217
900# Trip Throttle Valve Top Mechanism - 900# inlet and
900# outlet with Hard Packing, Oil Trip, Double Leak-off
07/16/1980
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Switches, Hand Relatch & Strainer, Mech. Trip Special Yoke
& Screw Spindle
Miscellaneous
MOV CIC: RCIC Trip and Throttle Valve Reset -RCIC-
MOV-MO14
06/15/2023
Miscellaneous
CIC: RCIC-MOV-
MO131
IT Basis Document - RCIC Motor Operated Steam Supply
Block Valve for RCIC Turbine
Miscellaneous
CIC: RCIC-MOV-
MO14
IST Basis Document - RCIC Turbine Trip Throttle Valve
Miscellaneous
COR002-18-02
Student Text - OPS Reactor Core Isolation Cooling
Miscellaneous
COR002-18-02R
Power Point Presentation - OPS Reactor Core Isolation
Cooling Requal
Miscellaneous
CR-CNS-2023-
2885-CA-001
Condition Potentially Affecting RCIC Steam Supply to RCIC
Turbine RCIC-MO-MO131
Miscellaneous
MOV CIC: RCIC-
MOV-MO131
Valve Information
06/15/2023
Miscellaneous
NEDC 91-185
MOV Thermal Overload Heater Sizing
Miscellaneous
NEDC 91-191
DC Equipment and Cable Short Circuit Withstand Ratings
Miscellaneous
RCIC-MOV-
MO131
Diagnostic Anomaly Trending
01/07/2020
Miscellaneous
VM-0023
CRD Hydraulic Control Units 729E950G1 729E950G2
29E950G3 729E950G4 729E950G5 & 729E950G6
Miscellaneous
VM-0903
Victoreen-Fluke Biomedical High Range Containment
Monitor 875
Miscellaneous
VM-0986
Limitorque Composite Manuals
Miscellaneous
VM-1750
GE Relay Composite Manual
Miscellaneous
VM-1778
ITT/American Heat Exchangers Composite Manual
Miscellaneous
VM-2021
Automatic Valve Composite Manual
Procedures
3.10
Flow Accelerated Corrosion (FAC) and Microbiologically
Influenced Corrosion (MIC) Program Implementation
Procedures
6.CRD.301
Withdrawn Control Rod Operability IST Test
Procedures
6.PRM.321
Containment High Range Area Monitor Functional Test
Procedures
6.PRM.322
Containment High Range Area Monitor Channel Calibration
Procedures
6.PRM.323
High Range Containment Area Monitor Victoreen 875
Source Calibration Check
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
6.PRM.329
Containment High Range Monitors A/R/H Determination
Procedures
6.RCIC.102
Operations Manual -Surveillance Procedure 6.RCIC.102
06/08/2021
Procedures
6.RCIC.102
Operations Manual -Surveillance Procedure 6.RCIC.102
06/09/2022
Procedures
6.RCIC.102
Operations Manual -Surveillance Procedure 6.RCIC.102
03/09/2023
Procedures
7.2.42.3
Heat Exchanger Tube Plugging
Procedures
EPIP 5.7.1
Emergency Classification
Procedures
SP 18-01
RCIC Turbine Control Upgrade Test - Order Number:
4742223
04/16/2019
Procedures
System Operating
Procedure 2.2.67
Reactor Core Isolation Cooling System
10/18/2022
Procedures
System Operating
Procedure
2.2.67.1
Reactor Core Isolation Cooling System Operations
07/13/2022
Work Orders
WO 287229, 5359111
Corrective Action
Documents
CR-CNS-
22-03404, 2022-03892, 2022-05890, 2023-02616
Drawings
DWG 3062,
Sheet 3
D.C. Control Elementary Diagrams
Drawings
DWG 729E261BC
HPCI System Piping and Instrumentation Diagram
Drawings
DWG 992C768
Outline (Induction Motor)
Drawings
DWG C-895-X,
Sheet 3
Motor and Solenoid Wiring Diagram
Drawings
DWG E150,
Sheet 14
Relay Settings for Battery Chargers & RPS MG Set Relays
Miscellaneous
Design Equivalent
Change Package
5476432
Residual Heat Removal (RHR) Pump Motor C Drain Plug
Replacement
Miscellaneous
Design Equivalent
Change Package
TCC-5501351
250 VDC Battery 1A Jumper
Miscellaneous
Equivalent
Replacement for Pressure Switch CNS-2-HPCI-PS-2787
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Change 4947302
Miscellaneous
VM 1188
25 & 250 Volt Batteries & Chargers
Miscellaneous
VM 1392
Square D Composite Manual
Miscellaneous
VM 1701
General Electric Motors
Procedures
2.1.12
Control Room Data
144
Procedures
2.2.24.1
250 VDC Electrical System (DIV 1)
Procedures
3-CNS-DC-112
Engineering Change Request and Project Initiation Process
Procedures
3-EN-DC-115
Engineering Change Process
15C17
Procedures
6.1EE.602
DIV 1 125V/250V Station Battery 92 Day Check
Procedures
6.EE.601
25V/250V Station and Diesel Fire Pump Battery 7 Day
Check
Work Orders
WO 5393705, 5458684
Corrective Action
Documents
CR-CNS-
21-00623, 2021-04536, 2022-00648, 2022-05246, 2023-
01885, 2023-01940, 2023-01947, 2023-01975, 2023-01978,
23-01994, 2023-02020, 2023-02507, 2023-02661
Drawings
DWG 2040,
Sheet 1
Flow Diagram Residual Heat Removal System
Drawings
DWG 2331-6-4
Pump Detail
Miscellaneous
VM-0004
16X20X28 1 Stage Civic Pump (Pump No. 280005/8) - RHR
Miscellaneous
VM-0008
Standby Liquid Control Pumps
Miscellaneous
VM-0023
CRD Hydraulic Control Units 729E950G1 729E950G2
29E950G3 729E950G4 729E950G5 & 729E950G6
Procedures
6.1DG.101
Diesel Generator 31 Day Operability Test (IST) (DIV 1)
Procedures
6.1OG.701
Augmented Off-Gas Hydrogen Monitor Channel Functional
(DIV 1)
Procedures
6.1RHR.101
RHR Test Mode Surveillance Operation (IST) (DIV 1)
Procedures
6.1SGT.501
SGT A Carbon Sample, Carbon Adsorber and HEPA Filter
In-Place Leak Test, and Components Leak Test (DIV 1)
Procedures
6.1SGT.501
SGT A Carbon Sample, Carbon Absorber and HEPA Filter
In-Place Leak Test, and Components Leak Test (DIV 1)
Procedures
6.2DG.101
Diesel Generator 31 Day Operability Test (IST) (DIV 2)
Procedures
6.2RHR.101
RHR Test Mode Surveillance Operation (IST) (DIV 2)
Procedures
6.PRM.322
Containment High Range Area Monitor Channel Calibration
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
6.PRM.323
High Range Containment Area Monitor Victoreen 875
Source Calibration Check
Procedures
6.SLC.101
SLC Pump Operability Test
Procedures
7.2.65
HCU Valve (111) Cartridge Replacement
Procedures
7.3.31.3
25V/250V Battery Terminal Cleaning and Torquing
Work Orders
WO 5394293, 5394309, 5394310, 5395872, 5395876, 5400535,
5441279, 5450290, 5454371, 5497725, 5501349
Miscellaneous
ANS Design
Report
A Prompt Alert and Notification System Design Report for
the Cooper Nuclear Station
Procedures
EPDG 2,
C-1
Semi-Monthly Alert and Notification System Siren Testing
Procedures
EPDG 2,
C-5
Annual Full-Cycle Sounding of Alert and Notification System
Procedures
EPIP 5.7.27
Alert and Notification System
Miscellaneous
ERO Call-In Test Results 2-25-2023
Miscellaneous
ERO Call-In Test Results 08-29-2022
Miscellaneous
Cooper Nuclear Station On-Shift Staffing Analysis
Miscellaneous
ERO Call-In Test Results 12-6-2021
Procedures
EPDG 2,
E-3
Quarterly ERO Call-In Test
Miscellaneous
50.54(q)
Evaluation
Number 2021-94
TSC/OSC Upgrade (Digital Upgrade)
01/10/2022
Miscellaneous
50.54(q)
Evaluation
Number 2022-22
CNS Emergency Plan, Revision 80
05/18/2022
Miscellaneous
50.54(q) Screen
Number 2021-94
TSC/OSC Upgrades (DEC-5416038)
01/07/2022
Miscellaneous
50.54(q) Screen
Number 2022-05
EPIP 5.7.20 Protective Action Recommendations Revision
03/18/2022
Miscellaneous
50.54(q) Screen
Number 2022-22
CNS Emergency Plan Revision 80
05/18/2022
Miscellaneous
50.54(q) Screen
Number 2022-30
CNS Emergency Plan Revision 81
06/20/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
50.54(q) Screen
Number 2022-45
Alert and Notification System Design Report, Revision 18
08/18/2022
Miscellaneous
50.54(q) Screen
Number 2022-45
Alert and Notification System Design Report
08/18/2022
Corrective Action
Documents
CR-CNS-
21-00925, 2021-01464, 2021-02561, 2021-03353, 2021-
03354, 2021-04900, 2021-05055, 2021-05300, 2021-05360,
21-05430, 2022-00028, 2022-00166, 2022-00443, 2022-
00470, 2022-00471, 2022-00472, 2022-00840, 2022-00863,
22-00926, 2022-01378, 2022-02040, 2022-02711, 2022-
2832, 2023-00278, 2023-00377, 2023-00596, 2023-01141,
23-01426, 2023-01613, 2023-01858,
Corrective Action
Documents
Resulting from
Inspection
CR-CNS-
23-02411, 2023-02414, 2023-02424, 2023-02437, 2023-
2471, 2023-02472
Miscellaneous
Drill Report: July 26, 2022, Full Team 4 & A Alternate
Facility Drill
07/28/2022
Miscellaneous
Drill Report: September 13, 2022, Full Team 3 & C Alternate
Facility Drill
09/20/2022
Miscellaneous
Drill Report: January 24, 2023, Full Team 3 & D Drill
01/30/2023
Miscellaneous
Drill Report: 2023 Medical Contaminated Individual Drill
03/22/2022
Miscellaneous
Drill Report: 2021 Radiological Monitoring Drill
2/20/2021
Miscellaneous
Drill Report: 2022 Medical/Contaminated Injured Person Drill
2/29/2022
Miscellaneous
Drill Report: March 11, 2021, Full Scale Team D Drill
04/21/2021
Miscellaneous
Drill Report: April 6, 2021, Biennial Dress Rehearsal
Exercise
05/05/2021
Miscellaneous
EPIP 5.7.21
Maintaining Emergency Preparedness - Emergency
Exercises, Drills, Tests, and Evaluations
Work Orders
5166232, 5377276, 5390636, 5412343
Corrective Action
Documents
CR-CNS-
23-01880, 2023-01899
Procedures
2.4OG
Off-Gas Abnormal
Procedures
5.1RAD
Building Radiation Trouble
Procedures
5.2FUEL
Fuel Failure
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
5.3DC125
Loss of 125 VDC
Procedures
5.7.1
Emergency Classification
Procedures
Secondary Containment Control (1-3)
71151
Miscellaneous
ANS Testing and PI Data 2Q2022 - 1Q2023
71151
Miscellaneous
DEP Opportunity PI Data 2Q2022 - 1Q2023
71151
Miscellaneous
ERO Drill Participation PI Data 2Q2022 - 1Q2023
71151
Procedures
0-EN-LI-114
Regulatory Performance Indicator Process
17C0
Corrective Action
Documents
CR-CNS-
21-02367, 2021-05170, 2022-00165, 2022-02221, 2022-
233, 2022-03577, 2022-05153, 2022-05941, 2023-01855
Miscellaneous
Design Equivalent
Change 5482558
Control Room AC Refrigerant Line Alternate Fittings
Procedures
0-CNS-WM-104
On-Line Schedule Risk Assessment
Procedures
0.39
Hot Work
Procedures
0.7.1
Control of Combustibles
Procedures
7.0.13
Control of Insulation Removal and Installation
Procedures
Radiation Worker Expectations
Work Orders
Corrective Action
Documents
CR-CNS-
22-02184, 2022-02874, 2022-06812, 2022-06814, 2022-
06897, 2023-00189, 2023-00401, 2023-01034, 2023-01057,
23-01066, 2023-01340, 2023-01366, 2023-01720
Drawings
DWG 2020
Flow Diagram Reactor Building Heating and Ventilation
Miscellaneous
Cooper Nuclear Station Engineering Desktop Guide 98-03-
Miscellaneous
11867624
Maintenance Plan Change Request
Miscellaneous
800000008838/9
Maintenance Plan
Miscellaneous
VM-0013
Model 7567F Relief & Safety Valves
Miscellaneous
VM-0440
Honeywell H&V Equipment
Miscellaneous
VM-1408
Honeywell Operators & Actuators
Procedures
0.40
Work Control Program
Procedures
2.2.47
HVAC Reactor Building
Procedures
3-CNS-DC-324
Preventive Maintenance Program
Work Orders