IR 05000341/2024002
| ML24220A158 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/14/2024 |
| From: | Hartman T NRC/RGN-III/DORS/RPB2 |
| To: | Peter Dietrich DTE Electric Company |
| References | |
| IR 2024002 | |
| Preceding documents: |
|
| Download: ML24220A158 (1) | |
Text
SUBJECT:
FERMI POWER PLANT, UNIT 2-INTEGRATED INSPECTION REPORT 05000341/2024002
Dear Peter Dietrich:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Fermi Power Plant, Unit 2. On July 17, 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.
Three findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
Licensee-identified violations which were determined to be of very low safety significance are 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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at Fermi Power Plant, Unit 2.
August 14, 2024
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 Fermi Power Plant, Unit 2.
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. 05000341 License No. NPF-43
Enclosure:
As stated
Inspection Report
Docket Number:
05000341 License Number:
NPF-43 Report Number:
05000341/2024002 Enterprise Identifier:
I2024002-0060 Licensee:
DTE Electric Company Facility:
Fermi Power Plant, Unit 2 Location:
Newport, MI Inspection Dates:
April 1, 2024, to June 30, 2024 Inspectors:
M. Domke, Senior Reactor Inspector J. Gewargis, Resident Inspector R. Ng, Senior Project Engineer T. Ospino, Resident Inspector J. Reed, Health Physicist T. Taylor, Senior 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 Fermi Power Plant, Unit 2, 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. Licensee-identified non-cited violations are documented in report sections: 71124.01 and 71152
List of Findings and Violations
Incorrect Temporary Battery Connection Leads to Fire Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000341/202400201 Open/Closed
[H.12] - Avoid Complacency 71111.24 A self-revealed Green finding was identified when the licensee failed to follow written instructions as required by MMA20, Work Execution and Closure, Revision 25. Specifically, the licensee failed to follow work order 66237704, which resulted in a fire and damage to electrical components in the balance-ofplant (i.e., non-safety-related) direct current (DC)distribution system.
Locked High Radiation Area Controls Not in Accordance with Technical Specifications Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000341/202400202 Open/Closed
[H.13] -
Consistent Process 71124.01 An NRC-identified finding of very low safety significance (Green) and an associated non-cited violation Technical Specification 5.7.2 was identified by inspectors when the licensee failed to adequately control access to an area having general area dose rates of up to 1200 mrem/hour at 30 cm from the radiation source.
Locked High Radiation Area Door was Unlocked and Unguarded Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000341/202400203 Open/Closed
[H.8] -
Procedure Adherence 71124.01 A self-revealed finding of very low safety significance (Green) and an associated violation of Technical Specification 5.7.2 was reviewed by inspectors when the door guard responsible for locked high radiation area (LHRA) access controls left their post unattended with the drywell access door unlocked.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000341/202300402 Access Control for Locked High Radiation Area 71124.01 Closed
PLANT STATUS
Fermi - 2 began the inspection period shutdown in Mode 5, plant cold shutdown, due to the ongoing refueling outage. The reactor achieved criticality following plant startup on May 9, 2024.
The main turbine generator was synchronized to the grid on May 12, ending the refueling outage. Power was raised to approximately 96 percent on May 14. Following this power ascension, reactor power was reduced to approximately 68 percent for a routine rod pattern adjustment. Subsequently, power was raised to 100 percent on May 16. On May 21, power was reduced to 94 percent for routine reactor recirculation pump motor generator surveillances. Power was restored to 100 percent on May 22. Power was reduced to 75 percent for a routine rod pattern adjustment on May 23. Power was then restored to 100 percent later, on May 23, and remained at or near 100 percent 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 onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Division 1 core spray during division 2 core spray simulated actuation surveillance test on June 11, 2024
- (2) Reactor core isolation cooling (RCIC) on June 20, 2024
- (3) High pressure coolant injection (HPCI) on June 21, 2024
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a full walkdown of the division 2 residual heat removal (RHR) completed the week of May 21, 2024.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 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) Drywell personnel access room during the week ending May 18, 2024
- (2) Auxiliary building cable tray room connected to relay room on June 12, 2024
- (3) RHR service water (RHRSW) complex division 1 service water pump room on June 13, 2024
- (4) Validation and walkdown of technical requirements manual (TRM) fire surveillance acceptance criteria on June 20, 2024
- (5) Check of reactor building main fire valves during the week ending June 29, 2024
- (6) Emergency diesel generator (EDG) 11 during the week ending June 29, 2024
71111.07 A - Heat Exchanger/Sink Performance Annual Review (IP Section 03.01)
The inspectors evaluated readiness and performance of:
71111.08 G - Inservice Inspection Activities (BWR) BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)
The inspectors evaluated boiling water reactor nondestructive testing, flaw evaluation, and welding activities by reviewing the following examinations from March 25, 2024, to April 4, 2024:
(1)
1. Volumetric examination by conventional ultrasound of ASME class 1,
category RA/R1.16, elements subject to intergranular or trans granular stress corrosion, reactor recirculation weld SWRS2B1-W1
2. Volumetric examination by conventional ultrasound of ASME class 1,
category RA/R1.20, elements subject to no degradation mechanism, reactor water cleanup weld FWG333096-9WF1
3. Surface examination by liquid penetrant on ASME class 1,
category BO/B14.10, pressure retaining welds in control rod drive housings, weld CRDH -X02Y31W2.
4. Technical evaluation TEE1122-064, Skipped Weld in division 2 RHR
heat exchanger support ring attachment weld
5. Pressure boundary weld FWR302182-0W205 under work order
no. 47458087 to replace service water return valve for EDG 13
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated training on primary and secondary containment technical specifications, emergency operation procedures, and emergency action levels on June 18, 2024.
71111.12 - Maintenance Effectiveness
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) Main steam isolation valve refurbishments prior to RF22 during the week ending May 4, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Issues with turbine bypass valve positions during startup from RF22 during the week ending May 11, 2024
- (2) Stator lift scheduled the week ending June 29, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Intermediate range monitor E restoration and past functionality on March 26, 2024
- (2) Agastat relays failing time delay requirement for low voltage pickup (multiple load shed relays) during the week ending April 20, 2024
- (3) HPCI valve E4150F042 seal-in contact issue (past operability) on April 26, 2024
- (4) LPCI valve E1150F017B failed asfound leak test on May 1, 2024
- (5) Mobile scaffolds in the reactor building basement not secured during the week ending June 29, 2024
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) Drywell moisture barrier replacement during the week ending April 20, 2024
- (2) General service water temporary modification for the refueling outage during the week ending April 20, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors continued their review of the refueling outage that was in progress when the assessment period began. The outage concluded on May 12, 2024. Major inspection activities involved containment closeout inspections, review of technical specification mode changes, and observation of refueling activities, reactor heatup, and reactor startup.
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) Recovery from balance-ofplant direct current (DC) battery bus voltage anomaly during maintenance, completed the week ending June 30, 2024
- (2) Emergency equipment (EE)SW/RHRSW piping replacement tie-in on April 12, 2024
- (3) LPCI valve E1150F015B following failed local leak rate test and repairs during the week ending May 11, 2024
- (4) Core spray valve E2100F006B following maintenance on May 15, 2024
- (5) LPCI valve E1150F017B restoration from failed asfound thrust test on May 17, 2024
- (6) Thru wall leak repairs from a drain line on the fuel pool cooling and cleanup system during the week ending June 15, 2024
- (7) EDG 12 following a maintenance outage on June 29, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) Drywell sand cushion inspections for license renewal during the week ending April 27, 2024
- (2) Loss of coolant/loss of offsite power testing of EDG 14 during the week ending May 4, 2024
- (3) Hydrostatic test of the reactor pressure vessel during the week ending May 4, 2024
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
- (1) Multi-facility emergency preparedness drill on May 28,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Workers utilizing the tool and equipment monitors and the whole-body contamination monitors while exiting the radiologically controlled area during the Unit 2 refuel outage
- (2) Radiation protection technicians performing surveys to release materials from contaminated areas around the drywell
- (3) Radiological labels for tools, equipment, and containers inside the radiologically controlled area
Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following radiological work:
- (1) High risk valve work
- (2) Control rod drive mechanism removal
- (3) Reactor core alteration work
- (4) Drywell moister barrier replacement
- (5) Under vessel work High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Locked high radiation area (LHRA) controls at S reactor water cleanup (RWCU) pump room RB213
- (2) LHRA controls for the drywell
- (3) HRA controls for room R24
- (4) LHRA controls for the fuel pool clean up heat exchanger and pump room
- (5) LHRA controls for the control rod drive mechanism drywell chute Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.03 - InPlant Airborne Radioactivity Control and Mitigation
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) HEPA unit H200005 used in the Drywell
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
OTHER ACTIVITIES-BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05)===
- (1) Unit 2 (April 1, 2023, through March 31, 2024)
MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) Unit 2 (April 1, 2023, through March 31, 2024)
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
- (1) Unit 2 (April 1, 2023, through March 31, 2024)
===71152 A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03) (2 Samples 1 Partial)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) A series of condition reports implying measurement and test equipment (M&TE)processes were not being followed, completed the week ending June 8, 2024
- (2) Identification that the RCIC containment isolation valve E5150F007 overall gear ratio did not match the design calculation, completed the week ending June 8, 2024 (3)
(Partial)
The inspectors continued their review of the information associated with the high pressure reactor pressure system (RPS) scram that occurred during the shutdown of plant for the refueling outage. Specifically, the inspectors reviewed the sites conclusions regarding the technical reason/mechanism that led to the high pressure scram.
71152 S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)===
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in that might be indicative of a more significant safety issue.
INSPECTION RESULTS
Incorrect Temporary Battery Connection Leads to Fire Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000341/202400201 Open/Closed
[H.12] - Avoid Complacency 71111.24 A self-revealed Green finding was identified when the licensee failed to follow written instructions as required by MMA20, Work Execution and Closure, Revision 25. Specifically, the licensee failed to follow work order 66237704, which resulted in a fire and damage to electrical components in the balance-ofplant (i.e., non-safety-related) direct current (DC)distribution system.
Description:
On April 1, 2024, workers were performing work order (WO) 66237704 to replace battery 2PC during the refueling outage. 2PC is a non-safety-related BOP 130/260 V direct current (DC)battery divided into two 130 V battery banks. Battery banks 2C1 and 2C2 each provide 130 VDC of opposite polarity. Connections are designed such that either polarity of 130 VDC, or 260 VDC, can be provided to the various DC loads in the system. Battery chargers 2C1 and 2C2 supply power to the associated 2C1 and 2C2 battery banks so that under normal operation, the batteries are maintained charged while the DC loads are powered. A spare charger, 2C12, is provided that can be connected to either battery bank.
During the replacement of the battery cells for 2PC, only one bank was going to be isolated and worked at a time, with power maintained to its associated loads via its normal charger and a temporary battery that would act to maintain the chargers voltage stable.
WO 66237704 provided steps to connect the temporary battery to the spare battery charger output, which would place the temporary battery onto the positive and neutral legs of the DC distribution system. This corresponded with the 2C1-half of battery 2PC. The WO also directed the output leads from battery 2C1 be lifted (so the battery could be disconnected from the system to facilitate cell replacement). While these steps were performed by the workers, they also removed the leads from spare battery charger 2C12 to the DC distribution system. This action was not directed by the WO, nor in the referenced plant procedure that aligned the batteries for performance testing (a configuration similar to the maintenance alignment being sought). With 2C12 no longer connected to the system, the temporary battery was also no longer connected to the system. As a result, normal charger 2C1 was on the distribution system without an associated battery, which caused excessive voltage swings to 130 V DC loads connected to the positive and neutral legs of the distribution system.
Shortly after the incorrect system alignment, alarms were received in systems powered from the associated legs of the DC system. Additionally, a fire was detected in the radiological waste building in the fuel pool cooling and cleanup demineralizer panel G41P010, which was promptly extinguished by an operator in conjunction with removing that circuit from service.
The next day, an acrid odor was detected in the condensate filter and demineralizer panel H21P250. After power was secured to that panel, degraded electrical components associated with the positive and neutral legs of the DC system were discovered.
Troubleshooting later revealed the oscillating voltages and the incorrect alignment of the DC system. Further investigation revealed other damaged electrical components, including power supplies in the dedicated shutdown panel. No operational plant transients occurred as a result of the electrical disturbance on the system.
Corrective Actions: The electrical configuration was corrected, the licensee investigated potentially affected loads throughout the DC system and performed troubleshooting/repairs of affected components.
Corrective Action References: CR202437731, and CR202437697
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to follow written instructions as required by MMA20, Work Execution and Closure, Revision 25, was a performance deficiency. MMA20 requires work groups to perform work in accordance with the instructions provided in the work package. Specifically, on April 1, 2024, the licensee failed to follow WO 66237704, Replace BOP 130/260 VDC Battery, which resulted in a fire and damage to electrical components in the balance-ofplant (i.e., non-safety-related)
DC distribution system.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to follow the WO steps led to instability in the BOP DC distribution system which damaged electrical components and started a small fire.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. Specifically, the inspectors utilized 1, Exhibit 2, Initiating Events Screening Questions, to assess significance.
Utilizing Section D, External Event Initiators, the inspectors determined the finding screened to Green, or very low safety significance, because the electrical components affected by the incorrect lineup could not result in a shutdown initiating event (as defined by IMC 0609 Appendix G, Shutdown Operations Significance Determination Process).
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the inspectors referenced the associated common language attribute QA.4 in NUREG2165, Safety Culture Common Language, when determining the cross-cutting aspect. QA.4 states that individual contributors perform a thorough review of the work site and planned activity every time work is performed, rather than relying on past successes and assumed conditions.
Further, individuals consider potential undesired consequences of their actions before performing work. In this case, action was taken outside the approved work order on the belief it would have no impact. However, the action resulted in an electrical disturbance which damaged plant components.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Locked High Radiation Area Controls Not in Accordance with Technical Specifications Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000341/202400202 Open/Closed
[H.13] -
Consistent Process 71124.01 An NRC identified finding of very low safety significance (Green) and an associated non-cited violation Technical Specification 5.7.2 was identified by inspectors when the licensee failed to adequately control access to an area having general area dose rates of up to 1200 mrem/hour at 30 cm from the radiation source.
Description:
On April 1, 2024, NRC inspectors were performing inspection activities related to high radiation area (HRA) and locked high radiation area (LHRA) controls. Radiological surveys of the drywell documented dose rates up to 1200 mrem/hour at 30 centimeters from the source of radiation. Observation of both drywell access control points identified scaffold structures had been erected adjacent to the drywell hatches with fencing installed to act as a barrier.
Flashing lights and LHRA postings were present on the scaffold fencing. Both scaffold structures had entry points approximately 3 feet wide with an electronic swing gate type turnstile installed to facilitate access to the area.
The licensee stated that control of the drywell access points in this manner was consistent with Technical Specification 5.7.3 which states: For individual areas accessible to individuals with radiation levels such that a major portion of the individuals body could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> dose > 1000 mrems with measurement made at 30 centimeters from the sources of radioactivity that are located within large areas such as reactor containment, where no enclosure exists for the purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be roped off and conspicuously posted, and a flashing light shall be activated as a warning device.
The licensee constructed an enclosure using a scaffold structure which could have included a locking gate to prevent unauthorized entry into the area. Consequently, the use of flashing lights is not permitted as described in Information Notice 8879 Misuse of Flashing Lights for High Radiation Area Controls. Therefore, drywell access to the LHRA should have been controlled via the requirements in 10 CFR 20.1601 (a)(3) or via Technical Specification 5.7.2.
Corrective Actions: LHRA door guards were posted at the access points to control entries into the LHRA.
Corrective Action References: CR202437903
Performance Assessment:
Performance Deficiency: The licensee failed to lock or provide continuous direct or electronic surveillance that is capable of preventing unauthorized entry to an LHRA.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to lock or guard access to the LHRA could have led to workers entering the area and receiving unintended dose.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (Green) because:
- (1) it did not involve aslow-as reasonably achievable planning or work controls,
- (2) there was no overexposure,
- (3) there was no substantial potential for an overexposure, and
- (4) the ability to assess dose was not compromised.
Cross-Cutting Aspect: H.13 - Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate. Specifically, the licensee implemented a new process for controlling access to the drywell LHRA without a systematic approach to make that decision.
Enforcement:
Violation: Technical Specification (TS) 5.7.2, High Radiation Area, requires, in part, areas accessible to individuals with radiation levels such that an individual could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a dose equivalent > 1000 mrems but < 500 rads at one meter from the sources of radioactivity shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift manager (SM) on duty and/or the radiation protection supervision. Doors shall remain locked except during periods of access by personnel under an approved radiation work permit (RWP) that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification or the RWP, direct or remote (such as closed-circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
Contrary to the above, from March 30 to April 4, 2024, the licensee failed to provide locked doors to prevent unauthorized entry, with the keys maintained under the administrative control of the SM or radiation protection supervision for areas accessible to individuals with radiation levels such that an individual could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a dose equivalent > 1000 mrems but < 500 rads at 1 meter from the sources of radioactivity. Specifically, both hatches to the drywell were controlled via electronic swing gate type turnstiles and flashing lights when locking gates could have been established.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Locked High Radiation Area Door was Unlocked and Unguarded Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000341/202400203 Open/Closed
[H.8] -
Procedure Adherence 71124.01 A self-revealed finding of very low safety significance (Green) and an associated violation of Technical Specification 5.7.2 was reviewed by inspectors when the door guard responsible for locked high radiation area (LHRA) access controls left their post unattended with the drywell access door unlocked.
Description:
On August 31, 2023, the drywell to Fermi Unit 2 was posted as an LHRA. Access control to the drywell LHRA was maintained by an LHRA door guard. Attachment 4 titled, LHRA ACCESS CONTROL GUARD RESPONSIBILITIES CHECKLIST, in Procedure 68.000.004, Radiological Posting and Labeling, was used to brief the door guard on their responsibilities for access control. The procedurally required briefing included instruction that the access control guard should remain stationed with a direct line-ofsight and control at the door until: a) access or barrier is secured or locked and verified by radiation protection (RP) or b) relieved by ANSI qualified RP personnel or c) relieved by another briefed access control guard. At approximately 1600, the LHRA door guard had a discussion with a foreign material exclusion zone guard. The result of this conversation led the LHRA door guard to leave the area unattended as they believed that their responsibilities as the control guard were no longer required. However, the drywell was still unsecured and as such required a door guard.
At approximately 1630, a relief for the LHRA door guard arrived at the drywell LHRA entrance and noted that there was no active door guard present and promptly notified RP and established control of the LHRA entry. RP then validated that no individuals were inside the drywell and that no unauthorized entry into the LHRA had been made while the door guard was not present.
Access to and work within HRAs, including LHRAs, need to be properly controlled to protect individuals from unplanned, uncontrolled exposures that could lead to overexposures. There are multiple options for controlling workers access to high radiation areas. One or more of these options must be used. A licensee can: use a control device to reduce radiation levels when a worker enters the area; or use an alarm to alert the worker and the supervisor of the activity when an entry is made; or keep the areas locked and maintain positive control over each individual entry; or use a person as a door guard that prevents unauthorized workers from entering the area while eliminating the condition of workers being locked inside the area.
Corrective Actions: The licensee revised their procedure to include additional instructions and controls when utilizing door guards for LHRA access control.
Corrective Action References: CR202332737
Performance Assessment:
Performance Deficiency: The licensee failed to lock or provide continuous direct or electronic surveillance that is capable of preventing unauthorized entry to an LHRA.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to lock or guard access to the LHRA could have led to workers entering the area and receiving unintended dose.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (Green) because:
- (1) it did not involve aslow-as reasonably achievable planning or work controls,
- (2) there was no overexposure,
- (3) there was no substantial potential for an overexposure, and
- (4) the ability to assess dose was not compromised.
Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, the door guard failed to follow procedure requirements for guarding a LHRA access point when they left their post without verification from RP staff.
Enforcement:
Violation: Technical Specification (TS) 5.7.2, High Radiation Area, requires, in part, that in addition to requirements of Specification 5.7.1, areas accessible to individuals with radiation levels such that an individual could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a dose equivalent > 1000 mrems but
< 500 rads at one meter from the sources of radioactivity shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift manager (SM) on duty and/or the RP supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification or the RWP, direct or remote (such as closed-circuit TV cameras) continuous surveillance may be made by personnel qualified in RP procedures to provide positive exposure control over the activities being performed within the area.
Contrary to the above, on August 31, 2023, from 1600 to 1630, the licensee failed to ensure that doors shall remain locked except during periods of access by personnel for areas accessible to individuals with radiation levels such that an individual could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a dose equivalent > 1000 mrems but < 500 rads at one meter from the source.
Specifically, the drywell was left unguarded when an LHRA door guard left their post without locking the drywell or having a replacement for the continuous direct control via door guard.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
The disposition of this finding and associated violation closes URI: 05000341/2023004-02.
Unresolved Item (Closed)
Access Control for Locked High Radiation Area URI 05000341/202300402 71124.01
Description:
A finding/violation was identified as described above.
Corrective Action Reference(s): CR202332737 Licensee-Identified Non-Cited Violation 71124.01 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Technical Specification 5.7.1 states, in part, pursuant to 10 CFR 20 paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area (HRA),as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as an HRA and entrance thereto shall be controlled by requiring issuance of a radiation work permit (RWP). Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
1. A radiation monitoring device that continuously indicates the radiation dose rate
in the area.
2. A radiation monitoring device that continuously integrates the radiation dose rate in
the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
3. An individual qualified in RP procedures with a radiation dose rate monitoring device,
who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the RP supervisor in the RWP.
Contrary to the above, on March 24, 2024, the licensee failed to ensure that an individual permitted to enter an HRA, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem/hr but < 1000 mrem/hr was provided a radiation monitoring device that continuously indicates or integrates dose rates or was accompanied by an individual qualified in RP procedures who is responsible for providing positive control over the activities.
Specifically, a worker entered the reactor cavity, an HRA, without their electronic dosimeter and without a RP technician escort.
Significance/Severity: Green. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609 Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green)because:
- (1) it did not involve aslow-as reasonably achievable planning or work controls,
- (2) there was no overexposure,
- (3) there was no substantial potential for an overexposure, and
- (4) the ability to assess dose was not compromised.
Corrective Action References: CR202437233 Licensee-Identified Non-Cited Violation 71152 A This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires, in part, that measures shall be established to assure that tools, gauges, instruments, and other measuring and test devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Licensee procedure MMA04, Measuring and Test Equipment Program, Revision 17, describes the process for controlling and maintaining measuring and test equipment (M&TE). Section 5.5 of MMA04 requires, in part, that if M&TE is not received, nor its location determined, by 30 days after its calibration due date, to declare the M&TE lost.
Section 5.7.1 requires corrective action documents to be written, with investigations documented, for out-ofcalibration or lost M&TE.
Contrary to the above, since 2021, the licensee failed to assure that tools, gauges, instruments, and other measuring and test devices used in activities affecting quality were properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Specifically, approximately 70 pieces of M&TE were not declared lost within 30 days of their calibration due dates after not being received, nor were their locations determined, going back to the year 2021. As a result, condition reports were not written, nor investigations performed as to the potential impacts on safety-related equipment. Further, there were several instances of M&TE that failed calibration in the year 2022 that did not have corrective action documents written nor investigations performed on potential impacts to safety-related equipment.
Significance/Severity: Green. The inspectors determined the issue was of very low safety significance due to answering no to the questions in Exhibit 2 of IMC 0609 Appendix A, The Significance Determination Process for Findings AtPower.
Corrective Action References: CR 202437005 Licensee-Identified Non-Cited Violation 71152 A This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR 50 Appendix B, Criterion III, Design Control requires, in part, that the applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Design calculation DC5719 Volume 1, Minimum Required Target Thrust (MRTT) for Generic Letter 8910 Gate, Globe, and Quarter Turn Valves (torque), and its associated design calculations document overall gear ratios (OARs) for safety-related valves. The OARs are a direct input to calculating motor operated valve (MOV) actuator output torque capability when testing the MOVs to ensure that under design basis conditions, the MOVs will perform their safety functions.
Contrary to the above, since original construction, the licensee failed to assure that the applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the design basis for several safety-related valves were not correctly translated into specifications.
The documented OARs in DC5719 Volume 1 did not match the OARs of the installed valves in the plant. The actual OARs were less than what was documented, resulting in non-conservative impacts to the calculated actuator output torques. Valves P4400F602A and B, the Emergency Equipment Cooling Water (EECW) makeup water tank outlet isolation MOVs for division 1 and 2, respectfully, had documented OARs of 52.0 in the design calculation (actual installed OARs were 36.5). Valve P4400F608, the division 2 EECW supply to drywell sump heat exchanger MOV, had a documented OAR of 25.38 in the design calculation versus the actual installed OAR of 13.6. Valve E5150F007, the reactor core isolation cooling steam supply inboard containment isolation valve, had a documented OAR of 49.0 in the design calculation versus the asinstalled OAR of 24.8.
As of the end of the assessment period, the licensee modified valve E5150F007 to change the OAR to what was documented in the design calculation. Condition reports were written for the remaining valves which contained actions to update the design calculation.
Significance/Severity: Green. The inspectors determined the issue was of very low safety significance due to answering yes to the first question in Exhibit 2 in IMC 0609 Appendix A, The Significance Determination Process for Findings AtPower.
Corrective Action References: CR 202438279 Observation: Storage and Staging Issues in the Reactor Building 71152 S Throughout the first two quarters of 2024, the inspectors identified several instances of material stored/staged in the reactor building that was not in compliance with station procedures (namely MOP23, Plant Storage, and MMA08, Scaffolding). Some examples are provided below:
- MOP23 requires a 3-foot standoff distance between equipment laydown areas and plant piping. The inspectors identified a laydown area with a storage box within 3 feet of sample lines for the division 2 residual heat removal service water radiation monitor. When notified, the licensee generated a condition report and moved the equipment. Several days later, the inspectors noted different material had been staged within 3 feet of the same sample lines.
- MOP23 requires a 2-foot standoff distance from important-tosafety equipment for certain mobile commodities. The inspectors identified a temporary power cart within 2 feet of an important-tosafety conduit (indication for alternate rod insertion and marked with orange paint, indicating division 1 equipment). When notified, the licensee generated a condition report. A few days later, the inspectors checked the area again and noted the cart had been moved, but only a few inches. The cart remained noncompliant with MOP23.
- The inspectors identified two unattended mobile scaffolds near safety-related instrumentation that were not restrained, nor had all the wheels locked, contrary to MMA08. The approved usage locations for the mobile scaffolds were not indicated on the scaffold tags, contrary to MMA08. When notified, the licensee generated a condition report (CR) and sent someone down to investigate. The next day, the inspectors found the wheels had been locked and scaffolds restrained, however, the restraints were applied in a manner that would not protect the nearby safety-related equipment (later that day the conditions were addressed when notified by the inspectors).
Throughout the period, the inspectors also noted the sites quality assurance organization identified numerous instances of improperly stored equipment.
The examples represent an adverse trend in the storage and staging of equipment in the plant and in establishing effective corrective actions, both near and long term. After a review of the individual issues, the inspectors concluded they were not of more than minor significance.
The licensee wrote CRs for the individual issues and acknowledged the trend.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 17, 2024, the inspectors presented the integrated inspection results to P. Dietrich, Senior VP and CNO, and other members of the licensee staff.
- On April 4, 2024, the inspectors presented the inservice inspection results to P. Dietrich, Senior VP and CNO, and other members of the licensee staff.
- On April 5, 2024, the inspectors presented the radiation protection inspection results to P. Dietrich, Senior Vice President and CNO, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202438853
RF22 AL Thrust Testing of E1150F017B Found Cracked
Anti-Rotation Device Tack Welds After Weld Repair
04/26/2024
CR202438907
RHR Pump B Vibration Reading is in the Alert Range
04/27/2024
CR202439143
E5150F007 Exceeded Maximum Allowable Torque Switch
Setting in CECO
05/02/2024
CR202439199
Inadvertent Start of RHR Pump B
05/03/2024
CR202439311
RCIC SPLY TO CST TEST ISO MOV E5150F022 Valve
Not Stroking Open
05/07/2024
CR202439525
HPCI Steam Supply Drain Pot Outboard Isolation Valve
Blowing Steam from Stem
05/12/2024
CR202439662
Pipe Cap Downstream of E4100F173/F174 Weeping
05/16/2024
CR202439896
WO Request-Hot Torque E4150F001 Following
05/29/2024
CR202440044
M&TE: Assumed Failure of MM198M, HEISE, PTE1/XT,
Handheld Calibrator Due to NOT Being Returned and Being
Days Past Calibration Due Date
06/04/2024
CR202440341
Repeat ORing Failures in HPCI Drain Pot E41N014 due to
Material Not Rated for Operating Temperature
06/17/2024
Corrective Action
Documents
CR202440343
E5150F005 EQ Moisture Seal
06/17/2024
6I7212211-07
Schematic Diagram Core Spray Inboard Isolation Valves A
and B E2150F005A and F005B
Q
6M7212034
Diagram Core Spray System (CSS) Reactor Building
6M7213144-1
Piping Isometric-North Core Spray Pump Discharge to
Reactor Pressure Valve (RPV) Penetration Reactor Building
Z
6M7215706-1
RHR Division 2 Functional Operating Sketch
M57081
HPCI System Functional Operating Sketch
Drawings
M57091
Reactor Core Isolation Cooling System Sketch Functional
Operating Sketch
Procedures
24.203.02
Division 1 CSS Pump and Valve Operability and Automatic
Actuation
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
24.203.03
Division 2 CSS Pump and Valve Operability and Automatic
Actuation
1400559
Alarm Bell Not Ringing
09/17/2023
29426
MAS511B Did Not Work When Tested
09/09/2023
29087
Fire Detector Base Lamp Did Not Illuminate
09/04/2023
CR202333258
South Entrance Door to Main Control Room Failed Fire
Door Inspection
09/22/2023
CR202333270
Investigation Needed for Fire Door to Sill Plate Gap
Acceptance Criteria
09/22/2023
CR202334331
Fire Door RA112 (Entrance to RW) Failed Door Gap
Inspection IAW 28.507.03
11/09/2023
CR202334423
RHR Fire Detector X82N504A Base Lamp Failed to Come
on During 28.505.51
11/14/2023
CR202334716
Door RA27 Inspection Failed Acceptance Criteria
per 28.507.03
11/28/2023
CR202335261
Door R36 (Div. 1 Battery Room) Acceptance Criteria Not
Met for 28.507.02
2/25/2023
Corrective Action
Documents
CR202439928
28.501.04 Past Critical for T8000F037
05/30/2024
Drawings
6M7215733-1
Fire Protection Functional Operating Sketch
CD
FPEE040009
Reactor Building First Floor Drywell Access and
Valve Room
FPEE090004
Requirement for Temporary Intervening Combustibles in
Modes 1, 2, and 3
Engineering
Evaluations
FPEE220009
Fire Door Gap and Repair Criteria
FPAB16 a
Auxiliary Building Cable Tray Area, North, Zone 6,
EL. 583'6"
FPAB16B
Auxiliary Building Cable Entry Room, Zone 6, EL. 583' 6"
FPAB16c
Auxiliary Building Cable Tray Area, South, Zone 6,
EL. 583'6"
FPAB29C
Auxiliary Building Cable Tunnel, Zone 9, EL. 613' 6"
FPRHR111-
RHR Complex, EDG 11 Room EL. 590' 0"
Fire Plans
FPRHR150
RHR Complex, Div 1 Pump Room, Zone 50, EL. 590' 0"
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
DC4921,
Appendix R
Compliance
Reactor Building First Floor Valve Room
E
28.507.02
Fire Door Surveillance Test
28.507.02
Fire Door Surveillance Test
28.507.03
Fire Door Inspection-BOP
Procedures
35.000.243
Repair/Replacement of Doors, Frames, and
Associated Hardware
Work Orders
298217
Perform 28.507.02 Fire Door
Inspection-Supervisory-Section 5.1 and
Att. 1LICENSE RENEWAL
07/26/2023
20381
Test Uncertainty Acceptance Criteria Not Met for 47.205.02
01/15/2021
CR202437729
RHR HX B Contingent Weld Repair WO Missed Reviews
04/03/2024
Corrective Action
Documents
CR202437790
RHR HX-After Blasting Inspection-Indications
04/03/2024
Fermi Unit 2 RF22
RHR B 9E1101B001B)
Miscellaneous
Heat Exchanger
Inspection Report
Division 2 RHR Heat Exchanger
04/2024
Procedures
47.205.02
RHR Division 2 (South) Heat Exchanger Performance Test
Corrective Action
Documents
CR202437688
RF22 IVUT: Relevant Indication Identified at Core Shroud
H3 Weld
04/01/2024
Corrective Action
Documents
Resulting from
Inspection
CR202437901
24 NRC License Renewal and Inservice Inspection
Observation
04/05/2024
Drawings
2182A
8-inch CL 150 OSY Gate Valve Manual Drive Weld End
Engineering
Evaluations
TEE1122-064
Skipped Weld in Div 2 RHR Heat Exchanger Support Ring
Miscellaneous
4701277857
Purchase Order for ASME Valve Material ID100322505
03/03/2020
RF22PT-002
Liquid Penetrant Examination
04/03/2024
RF22UT-004
UT Calibration/Examination
03/31/2024
NDE Reports
RF22UT-005
UT Calibration/Examination
03/30/2024
71111.08 G
Work Orders
47458087
Weld Process Control Sheet
04/10/2022
Miscellaneous
LPOP2022414
License Operator Requalification Training:
01/11/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
RF22 Modifications
CR202436928
RIT Identified Incorrectly Supplied/Identified Parts for
Main Steam Isolation Valve Rebuild
03/14/2024
CR202437014
RF22 Valve Team-Broken Piston Ring When Rebuilding
MSIV Actuator
03/19/2024
CR202437026
RF22 Valve Team-Main Steam Isolation Valve Rebuild
Parts Failure, Rod Packing
03/20/2024
CR202437109
B2103F022C and B2103F028C Hydraulic Pistons
Accidentally Swapped
03/22/2024
CR202437330
Inboard MSIV B2103F022D Unqualified Coating Not
Removed Prior to Assembly Under WO 69932883
03/26/2024
CR202437736
RF22 Valve Team Issue Encountered-WO 69787880
04/02/2024
Corrective Action
Documents
CR202437789
Unqualified Coating on MSIV Actuator
Serial Number 81821091
04/03/2024
6M7216141-1
26" Main Steam Isolation Valve CYL Operated 21" Diameter
Seat Bore Sheet 2 Misc. Details, Sections and Notes
C
Drawings
6M7216141-4
26" Main Steam Isolation Valve Cycle Operated
21" Diameter Seat Bore Sheet 5, 20" Pneumatic Actuator
and 6" Hydraulic Actuator Details
E
Miscellaneous
MVR342.4
Key Safety-Related Component EQ Component
Procedures
HEP304
Engineering Procedure
Work Orders
69787880
Shop Work-Refurbishment of MSIV B2103F022D
Actuator and Manifold
11/10/2023
Corrective Action
Documents
CR202439431
Main Steam West Bypass Valve Not Operating as Expected
05/10/2024
193823 51 1000
Fermi 2 Evaluation of Heavy Loads Impact During
Main Generator Replacement
IPTE 2401
MOP010223 IPTE
2401
Infrequently Performed Test or Evolution Brief Sheet
IPTE 2401
Engineering
Evaluations
TEN3022-041
Generator Haul Path Evaluation
Miscellaneous
IPTE 2401
IPTE 2401 IPTE Brief Valve List
Work Orders
68129047
EDP80000 Transport New Stator to TB1
05/30/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
68129189
EDP80000 Transport New Stator to TB3 Staging Area
07/23/2023
Calculations
DC5719
Volume I
MRTT for Generic Letter 8910 Gate, Globe, and Quarter
Turn Valves (Torque)
Calibration
Records
Division 1 IRMs
Time Domain
Reflectometry
Division 1 IRMs Time Domain Reflectometry
0011142
Valve Failed to Auto Close During PMT
01/26/2000
0310957
E4150F003 Not Stroking Closed
21564
Valve Not Stroking Closed (E4150F003)
05/06/2004
21572
Evaluate Primary Containment Isolation Signal Seal-In on
the E4150F003
24169
HPCI Torus Suction Inboard Isolation Valve E4150F042
Remote Manual Open Seal-In Not Functioning
05/18/2012
28443
Request Trending for Main Contactor Aux Contact
Operation
11/21/2014
24894
Continued Margin Tracking: Low Torque/Thrust Margin in
Multiple MOVs
06/15/2016
25849
Margin Improvement Required for E1150F017B
07/21/2017
27001
AsFound Test on Open Contactor
08/21/2017
27028
Low Coil Pickup Voltage
08/22/2017
2033151
TSR 38064 Improperly Evaluated DC Relay Minimum
Required Voltage
2/22/2020
28799
TR Relay Did Not Meet Pickup Acceptance Criteria
10/28/2021
9711485
MCC ITE Series 5600 Replacement Buckets
CR202435667
IRM E (C51R607C) Indicating High
01/18/2024
CR202436518
E4150F042 HPCI Booster Pump Suction from
Suppression Pool Isolation MOV Failed to Close
2/25/2024
CR202436518
E4150F042 HPCI Booster Pump Suction from
Suppression Pool Isolation MOV
2/25/2024
CR202436970
HPCI Suppression Pool Inboard Isolation Valve E4150F042
Potentially Not Meeting Required Thrust into Closed Seat
03/17/2024
CR202437187
IRM E Reading High out of Tech Spec Limit for Mode 4
03/23/2024
Corrective Action
Documents
CR202437387
E1150F017B Thrust Test Did Not Meet Acceptance Criteria
03/27/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
for As-Found Available Closing Thrust
CR202437416
Good Catch Prevented Possible Overthrust on
E1150F017B
03/28/2024
CR202437812
X51 Overcurrent Relay Failed AsFound Timing Test
Acceptance Criteria and Could Not Be Adjusted to Pass
AsLeft Timing Test Acceptance Criteria
04/04/2024
CR202437875
QA1 DC Relay Failing Degraded Voltage Pick-Up Testing
04/05/2024
CR202438077
DC6480 Vol 1 Requires Revision for Agastat 7000 Series
Time Delay Relays
04/11/2024
CR202438143
Relays 1MH62 1LV62 and 1LU62 Failed AsFound
Acceptance Criteria
04/15/2024
CR202439129
Suspect I/V and TDR Traces Discovered for IRM F
During Performance of 45.000.003
05/02/2024
20045
NRC ID Concern-Failed Relays During Bench Testing
2/09/2009
CR202440178
NRC Identified: Unrestrained Mobile Scaffolds in Reactor
Building Basement NW Quad
06/10/2024
Corrective Action
Documents
Resulting from
Inspection
CR202440363
NRC Identified: Missed Past Operability Review on
CR202440178
06/18/2024
6I7212221-08
Schematic Diagram HPCI Suppression Pool Isolation
Valves
Z
6I7212572-29
Schematic Diagram 4160V Ess. Buses 65E and 65F Load
Shedding Strings
O
Drawings
6SD7212530-14
Frontal Elevation 260VDC MCC 2PB1 Division 2 Auxiliary
Building 3rd Floor
Design Basis
Document A3100
Valves
07/19/2019
EFA-R16001
Pick-Up Voltage Acceptance Criteria for CR120B GE
Relays
ERE 31955
Replacement of Nebula EP with MOV Long Life Grease
11/08/2006
ERE 32092
MOV Lubrication Change
2/28/2008
ERE 45088
MOV Motor E115017A and E115017B Replacement
A
Engineering
Evaluations
TEE1124-026
E1150F017B Past Operability-MOV Failed AsFound
Diagnostic Thrust Testing
05/22/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
EPRI 100014
Circuit Breaker Maintenance Programmatic Considerations
EPRI 1013457
Nuclear Maintenance Applications Center: Switchgear and
Bus Maintenance Guide
2/01/2006
EPRI 1013463
Nuclear Maintenance Applications Center: Application
Guide for MOVs in Nuclear Power Plants
EPRI Report
NP6229R1
Technical Repair Guidelines for Limitorque Model SMB000
Valve Actuators
2/01/1994
Nuclear Maintenance Applications Center Lubrication Guide
Generic Letter
8910 Close Out
Generic Letter 8910 Close Out
08/01/1995
Letter to Scott
Shephard from
Bob Keck
Fermi 2 Testing Input of time Delay Agastat Relay Model
E7012P and Synchro-Start Speed Switch Model ESSB2AT
for Use in DC6480 Vol. 1 Rev. 0
2/12/2011
TR106563V1
EPRI Application Guide for MOVs in Nuclear Power Plants
09/01/1999
TSR/38064
Design Review
VMC224.3
LIMITORQUE Valve Actuators Vendor Manual
I
Miscellaneous
VME518
Vendor Manual: Spectrum Technologies Series 5600
Motor Control Center
03/31/2009
35.306.003
Limitorque Motor Operator-Periodic Inspection
35.306.018
Spectrum Technology Motor Control Center
Load Compartment
35.306.020
Motor Operated Valve Mini Periodic Inspection
47.306.01
Analysis of Motor Operated Valves
47.306.02
Votes System Operating Procedure
17A
47.306.06
MOV Diagnostic Testing with the Quiklook 3 System
MMA08
MOP23
Plant Storage
Procedures
PEP03
Motor Operated Valve Program
000Z944674
Perform Dynamic Valve of E1150F015B and E1150F017B
05/14/1994
27094013
Perform Mini Periodic MOV Inspection
11/02/2010
30084578
Perform MOV Thrust (VIPER) Testing Per GL 9605
Program
11/02/2010
Work Orders
36093716
E1150F017B-MOV Motor Inspection Using
10/04/2015
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Borescope/Videoprobe
3846513
Packing Leak on E1150F017B
10/04/2015
2942324
Perform Maxi Periodic MOV Inspection
04/18/2020
243974
Perform Mini Periodic MOV Inspection
03/29/2017
44484757
Inspect/Test 260 VDC MCC Bucket MCC 2PB110B
Feeds E4150F42
08/21/2017
46576056
Inspect/Test 65E Load Shed Strings AB Auxiliary Relays
and 65E-E6 Overcurrent Relays
11/27/2016
242687
Perform Mini Periodic MOV Inspection and MPM
03/30/2022
A535100100
Inspect/Test 260 VDC MCC Bucket MCC 2PB110B
Feeds E4150F042
03/02/2011
E365010100
Perform MOV Thrust Verification Testing
11/01/2001
CR202436951
Nuclear Oversight - Gaps in Readiness to Execute Temp
Mod-230010, Alternate Makeup to Ultimate Heat Sink
(UHS)
03/15/2024
CR202437041
RF22 Initial Coating Applicator Qualifications for
UT15 are UNSAT
03/22/2024
CR202437383
AsFound Moisture Barrier Inspection Reportable
Conditions
03/27/2024
Corrective Action
Documents
CR202437605
RF22 - Drywell Moisture Barrier Replacement - Average
Surface Profile Exceeded Specified Range AZ 180
to 359 Degrees
03/31/2024
6M7215720
Circulating Water System Functional Operating Sketch
BF
Drawings
6M7215726
General Service Water System Functional Operating Sketch
CG
Engineering
Changes
230010
Alternate Makeup to the Ultimate Heat Sink
Engineering
Evaluations
TEP4123-053
Evaluation of Temporary Supports for RBCCW HX ISO
Valves P4100F047A, P4100F047B, and P4100F049B
Repairs
Miscellaneous
EDP70368
Drywell Moisture Barrier Seal Replacement
[CONFIDENTIAL]
Procedures
43.000.019
Primary Containment Inspection
Work Orders
44743728
RBCCW TCV Outlet Valve is Very Difficult to Operate in the
08/16/2023
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Closed and Open Direction
2481797
P4100F064 GSW to Turbine LO Cooler Isolation Valve
Leaking By
03/12/2024
53820458
P4100F049 Leak Thru-East RBCCW HX Not Isolating
03/15/2024
56744350
Replace RBCCW HX Isolation Valve P4100F047B
2/22/2023
56744362
Repair/Replace GSW Valve P4100F049B
04/03/2024
61446224
Leak on P4100F840 GSW Pipe Seat Leakage on Valve
63391492
P4100F086 GSW TBCCW TEMP CRTL VLV P43F402 TCV
Outlet ISO VLV Could Not be Fully Closed
08/16/2023
64514521
Repair P4100F109 (GSW TO Cyclone Separator HDR ISO
VLV) LEAKING BY
05/08/2023
64514600
Valve P4100F806 Leaking By, Replace Valve
05/01/2023
260776
Replace P4100F064 GSW to Turbine Oil Coolers Isolation
09/05/2023
66506662
Replace GSW Piping in RF226" GSW Cooling Water
Supply Header for Circ Water Pumps Motors
2/16/2024
67715790
Install / Remove Temporary Station Air Supply to Support
RF22
08/16/2023
68491563
P4100F070 GSW to Turbine LO Coolers Isolation Valve
Leaking By
2/09/2024
68491682
P4100F072 GSW to Turbine LO Coolers Isolation Valve
Leaking By
06/01/2023
68491793
P4100F075 GSW to Turbine LO Coolers Isolation Valve
Leaking By
03/12/2024
68531512
Install / Remove Temporary Jockey Fire Pump in RF22
09/19/2023
69376671
Leak from GSW Header on TB1 East Aisle
CR202437844
Fatigue Management-10 CFR 26 Subpart I Work-Hour Rule
Violations
04/06/2024
CR202437963
NRC Feedback-Limited Details Provided in Electrical CR
(202437830)
04/06/2024
CR202438367
Relevant Indications in Recirc Small Bore Socket Welds
04/14/2024
Corrective Action
Documents
CR202438442
E4150F002 Thrust Testing Results Below 90 Percent
Correlation Limit-AF and AL LLRTs Needed
04/16/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202438481
Fatigue Assessment Being Performed
04/18/2024
CR202439028
1D78 RCIC Inlet Steam Drain Pot Level High Alarm due to
InLeakage
04/30/2024
CR202439050
RF22 RPV System Leakage Test Leakage Identified from
E4150F002
05/01/2024
CR202439299
Nuclear Oversight Potential Non-Conservative Assumption
in Crediting RWCU Heat Removal Capability as Alternative
Method of Decay Heat Removal
05/07/2024
Drywell Instrumentation Isometric Piping from B311L012B,
L015B, L016B, to Penetration X33A Reactor Building
Drywell
H
Drywell Instrumentation Isometric Piping from B31L010B,
B31L010A to Penetration X33A Reactor Building Drywell
Unit 2
H
6M7215706-1
RHR Division 2 Functional Operating Sketch
AN
6M7215706-1
RHR Division 2 Functional Operating Sketch
6M7215706-2
RHR Division 1 Functional Operating Sketch
AE
6M7215707
CSS Functional Operating Sketch
6M7215711-1
Reactor Water Clean Up Reactor Building Functional
Operating Sketch
6M7215712-1
Fuel Pool Cooling and Cleanup System
T
6M7215712-1
Fuel Pool Cooling and Cleanup system
Functional Operating Sketch
T
6M7215712-2
Fuel Pool Filter Demin System Functional Operating Sketch
K
Drawings
6SD7212500-01
One Line Diagram Plant 4160V and 480V System Service
RF22 Outage Nuclear Safety Review
01/09/2024 -
01/11/2024
Shutdown Safety
Risk Management
Planning Form
RF22 Reactor Cavity Flood-up Evolution
Miscellaneous
Shutdown Safety
Risk Management
Planning Form
RF22 Shutdown Cooling Out of Service for 24.404.03
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Shutdown Safety
Risk Management
Planning Form
RF22 General Service Water Outage
Shutdown Safety
Risk Management
Planning Form
RF22 Reactor Cavity Drain Down Evolution
Temporary
Change Notice -
13188
Pressure/Temperature Monitoring During Heatup and
Cooldown
09/05/2023
2.000.02
Plant Startup to 25 Percent Power
111
23.623
Reactor Manual Control System
43.401.500
Local Leakage Rate Testing for Penetration X7A, X7B,
X7C, and X7D
Procedures
MOP05200
31129 ESW 15 05
004
RHRSW and EESW Piping Replacement Contingency
Study-RHRSW Orifice Sizing During Refueling Outage
24
29190
E1150F017B
Calculated Survivable Thrust for E1150F017B
2/15/1993
Calculations
DC5038 Volume
XIV
Torque/Thrust Calculation for Valves E1150F017A and
E1150F017B
C
25849
Margin Improvement Required for E1150F017B
07/17/2017
CR202436841
License Renewal-Plant Configuration Does Not Match
Drawing M2219
03/13/2024
CR-202437390
43.401.516 Test 1 for E1150F015B Leakage Measurement
Error
03/27/2024
CR202437697
RF22 DC System Voltage Anomaly - Fire Inside of the
FPCCU Demin Panel G41P010
04/01/2024
CR202437731
RF22 DC System Voltage Anomaly-Investigate BOP
Battery Bus
04/02/2024
CR202437830
Hour Critical Path Loss Due to Challenges with
E1150Fo15B Maintenance
04/23/2024
CR202437832
Pressure Isolation Valve Test Frequency Change Required
04/04/2024
Corrective Action
Documents
CR202438008
E1150F015B Exceeded its Total Maximum Leakage
04/07/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202438067
Through Wall Pipe Leak Upstream of G4100F043 on RB3
04/09/2024
CR202438451
RF22 Critical Path Delay of 14 Hours due to E1150F015B
Maintenance Activity
04/16/2024
CR202438517
Cracked Disc Found in E1150F015B During
RF22 Inspection
04/18/2024
CR202438526
License Renewal AMP B.1.12 - Sand Accumulation Found
in 90 Azimuth Drywell Sand Cushion Drain Line
04/18/2024
CR202438527
License Renewal AMP B.1.12 - Potentially Degraded Pipe
Sections Found in 0/360 Azimuth Drywell Sand Cushion
Drain Line
04/18/2024
CR202438528
License Renewal AMP B.1.12 - Sand Accumulation Found
in 180 Azimuth Drywell Sand Cushion Drain Line
04/18/2024
CR202438529
License Renewal AMP B.1.12 - Sand Accumulation Found
in 270 Azimuth Drywell Sand Cushion Drain Line
04/18/2024
CR202438654
E1150F017B Motor Stall and Overthrust during
EDP 800007 Implementation
04/21/2024
CR202438740
E2100F006B, CSS2 Check Valve, Not Closed
04/23/2024
CR202438988
E2100F006B Failed to Stroke During
Surveillance 24.203.02
04/29/2024
CR202440039
Oil-Soaked Insulation/Fire Hazard
06/04/2024
CR202440042
Broken Lockwire Found on VD 8 Bolt Flange
06/05/2024
CR202440045
EDG 12 Lower Crankshaft Seal Oil Leak
06/04/2024
CR202440059
EDG12 SSO-Small Bore Line Entering Turbocharger
Cooling Water Line Interference Resolution Document
Control Disposition
06/06/2024
CR-202440075
Crushed Air Line
06/06/2024
CR202440084
Cracked Reducing Bushing on EDG 12
06/06/2024
CR202440124
DPM Leak Idd on EDG 12 Turbo Charger Coolant Line
06/07/2024
103262
24" 600 Pound Angle Key-Bushing E115017B
M
106903
20" 900 Pound Y Globe Yokearm
05/02/1972
6C7212358
Reactor Building and Auxiliary Building Framing
Section 4040 Lower
T
Drawings
6C7212407
Reactor Building Framing Sections and Details
Q
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
6I7212735-02
Loop Diagram Dedicated Shutdown Panel H21P623
Instrumentation
J
6I721N271117
Schematic Diagram Emergency Diesel Generator 11, 12,
13, and 14 Exciter - Voltage Regulator
S
6M7212083
Diagram RHR Division II
6M7212219
Floor and Equipment Drains Basement and First Floor
Reactor Building
6M7215357
EECW System Division II
04/15/2024
6M7215706-1
RHR Division 2 Functional Operating Sketch
AN
6M7215706-3
RHRSW Makeup Decant and Overflow Systems Functional
Operating Sketch
6M7215729-2
EECW (Division 2) Functional Operating Sketch
6M7215734
Functional Operating Sketch Emergency Diesel Generator
System
6SD7212530-12
One Line Diagram 260/130 BOP Battery 2PC Distribution
BB
7M7212053
Reactor Building Piping Plan Basement (West) EL. 562' 0"
I220002
Logic Diagram RHR
E
I220119
Schematic Diagram RHR Loop B to Recirculation Outboard
Isolation Valve E1150 F017B
Q
80029
Tie-Ins Buried Pipe Replacement for RHRSW and EESW
Systems
Engineering
Changes
EDP 800007
Torque Switch Bypass on MOV E1150F017B
04/09/2024
06054 ISI-NDE
Appendix J - Generic Letter 8910 Correlation (Retest
Guidelines for Appendix J Valves)
Engineering
Evaluations
TEE1124-021
Evaluate Motor Stall Impacts on RHR DIV 2 LPCI MOV
E115F017B
04/23/2024
DBD-C3600
Design Basis Document for Dedicated Shutdown System
E
DEC 70119
Replacement Packing End Wiper Rings in Valves and
Restrict the Usage of Stainless Steel Spacers
DEC 70119
Replacement Packing End Wiper Rings in Valves and
Restrict the Usage of Stainless Steel Packing Spacers
Miscellaneous
T2301A001
Drywell Sand Cushion Drain Line (90 Degree Azimuth)
04/17/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Temporary
Change Notice -
13188
Pressure/Temperature Monitoring During Heatup and
Cooldown
09/05/2023
TMPLAM
Pressure and Temperature Limits Report [CONFIDENTIAL -
PROPRIETARY INFORMATION]
VMC216
Valves Various William Powell Co.
J
23.208
RHR Complex Service Water Systems
135
24.137.21
Reactor Pressure Vessel System Leakage Test
24.208.03
Division II EESW and EESW Makeup Pump and Valve
Operability Test
24.307.04
Emergency Diesel Generator 14 - Loss of Offsite Power and
ECCS Start with Loss of Offsite Power Test
35.000.230
Pressure Seal Valve General Maintenance
35.306.001
480 Volt Switchgear Breaker and Relay Control Testing
37.000.018
60Month BOP 130/260 Battery Capacity Test
MES23
Inservice Inspection and Testing
23.1
MES28
Leakage Reduction and Primary Containment Leakage
Rate Programs
MES46
ASME Section XI Containment Inservice Inspection
Program
8.2
Procedures
SOE 2201
Post-Maintenance Testing and Run for Division II RHRSW,
EDGSW, and EESW in RF22 Post Piping Replacement
Modification
04/11/2024
64556970
Change EDG 12 Governor/Booster Oil
2/27/2022
64658009
03/02/2022
237704
Replace BOP 130/260 VDC Battery
04/01/2024
66492336
Perform 24.307.04 EDG 14 - Loss of Offsite Power and
ECCS Start with Loss of Offsite Power Test
05/01/2024
66675382
66675382 Thrust Trace/Curve AsFound (Test 0)
E1150F017B
03/27/2024
Work Orders
66675382
IST Required Perform MOV Diagnostic Testing
E1150F017B
03/27/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
67887597
ECN80029 - DIV 2 - PHASE 6 - RF22 - RHRSW Orifice
Plate Replacement
01/09/2024
67887617
ECN 80029 DIV 2 Phase 6 RF22 PMT RHRSW Hydro
04/24/2024
67887619
ECN80029 DIV 2 Phase 6 RF22 PMT EESW Hydro
01/06/2024
68592829
Clean EECW Div. 2 B Plate Type Heat Exchanger
05/14/2024
68592830
Clean EECW Div. 2 D Plate Type Heat Exchanger
01/23/2024
70084722
Draining/Restoration for the EDG 12 SSO
11/13/2023
70647173
Perform SOE 2201
04/18/2024
262695
License Renewal - Plan Config. Sand Cushion Drain Line
Does Not Match Drawing M2219
04/17/2024
71495466
71495466 Thrust Trace/Curve Torque Switch Bypass Test 1
No Overthrust
04/21/2024
71495466
71495466 Thrust Trace/Curve Overthrust (Test 2)
04/21/2024
71495466
71495466 Thrust Trace/Curve No Overthrust AsLeft
(Test 3)
04/26/2024
71495466 Thrust
Data
E1150F017B App III PV and EDP 800007
04/26/2024
71590010
Dedicated Shutdown Panel H21P623 Does Not Have
Power. Replace C36K405 if Required
04/15/2024
71735194
E1150F015B Exceeded its Total Maximum Leakage
04/14/2024
71737617
Final 43.401.516 RHR Pressure ISO Valve Leakage
Test-1: E1150F015B
04/25/2024
71789624
Through Wall Pipe Leak Upstream of G4100F043 on RB3
04/08/2024
296017
WELD Repair E1150F017B Key-Bushing and Yoke
04/22/2024
2479093
E2100F006B CSS2 Check Valve Not Closed Troubleshoot
and Repair
04/26/2024
Corrective Action
Documents
CR202440117
Vaporstreams Training Environment Was Used During the
May 28th RERP Drill
06/07/2024
Miscellaneous
Technical Support
Center - OBJ
Evaluations
Drill Gold Team Drill/Green Team Controllers
20.300.PHASE
Loss of Phase
Procedures
EP101
Classification of Emergencies
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202332737
LHRA Door was Unsecured
08/31/2023
Corrective Action
Documents
CR202437233
Individual Entered LHRA Without Proper Dosimetry
03/24/2024
Corrective Action
Documents
Resulting from
Inspection
CR202437779
NRC Observed - RPT Responding to TEM Alarm Not
Wearing Gloves
04/03/2024
Engineering
Evaluations
NPRP230034
Status of Fermi 2 Alpha Source Term
67.000.511
Radiological Job Coverage for High Radiation Diving
Evolutions
68.000.002
Radiation and Contamination Surveys (NISP-RP02)
68.000.003
Radiological Air Sampling (NISP-RP03)
68.000.004
Radiological Posting and Labeling (NISP-RP04)
68.000.005
Access Controls for HRAs (NISP-RP05)
68.000.006
Personnel Contamination Monitoring (NISP-RP06)
68.000.007
Control of Radioactive Material (NISP-RP07)
Procedures
MRP04
Radiation Protection Conduct Manual
Air Sample for
Drywell Under
Vessel
Airborne Radioactivity Calculation Worksheet for Drywell
Under Vessel
04/02/2024
Air Sample for
RB5
Airborne Radioactivity Calculation Worksheet for Reactor
Building 5
03/24/2024
Air Sample for the
reactor cavity
Airborne Radioactivity Calculation Worksheet for the
Reactor Cavity
2/25/2024
Nuclear
Generation
Memorandum
0801.26
Status of Fermi 2 Alpha Source Term
04/26/2023
PM2024032342
RF22 Drywell Basement Initial Survey
03/23/2024
PM2024032426
Update Cavity Survey with RPV Head Installed
03/24/2024
PM202403249
Survey After Drywell Head Removal
03/24/2024
PM2024032623
RF22 DADO and DW Basement Updated Smear Survey
03/26/2024
Radiation
Surveys
PM2024040120
NW MSR Initial Survey
04/01/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
PM2024040141
Torus Diver Suit Survey
04/01/2024
PM202404024
Pre-Dive Survey (Bay 1-Bay 5)
04/02/2024
RWP 242024
Valve Group BOP Tasks
RWP 243015
RF22 Under Vessel Support
RWP 244002
RB5-Reactor Core Alterations, Bridge Maintenance, LPRM
Replacement and Support Activities
RWP 245002
E1150F068B Replacement Tasks
Radiation Work
Permits (RWPs)
RWP 245009
Drywell Moisture Barrier (Dado) Project
TEDE 243004
Respirator Evaluation Worksheet for under vessel work
03/20/2024
Calculations
TEDE 244005
Respirator Evaluation Worksheet for Reactor Cavity and
Dryer/Separator Pit
03/20/2024
Corrective Action
Documents
CR202437723
RHR HX Coating Project-HEPA Not Working-Had to
Switch Out
04/02/2024
65.000.541
Filter Leak Testing of Portable HEPA Ventilation Units and
HEPA Vacuums
65.000.704
Issuance of Respiratory Protection Equipment
Procedures
68.000.008
Use and Control of HEPA Filtration and Vacuum Equipment
(NISP-RP08)
MSPI Derivation Report-High Pressure Injection System
Unreliability Index
03/2024
MSPI Derivation Report-High Pressure Injection System
Unavailability Index
03/2024
MSPI Derivation Report-Emergency AC Power System
Unreliability Index
03/2024
MSPI Derivation Report-Emergency AC Power System
Unavailability Index
03/2024
Fermi 2-EDG MSPI and WANO Performance Indicators
Various
Miscellaneous
Fermi 2-HPCI Performance Indicators
Various
71151
Procedures
24.000.05
Eight Hour-Mode 1, 2, 3-Control Room RCS Operation
Leakage
Various
71152A
Corrective Action
Documents
23796
Alarm 4D91 Received During 24.109.02 Turbine Bypass
Valve Operability Test
07/08/2007
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
23183
Alarm 4D91 Was Received During 24.109.02 Turbine
Bypass Valve Operability Test
05/11/2008
24355
N1100F059A MS East Bypass Valve Discrepancy
During 24.1.09.02
05/17/2021
24355
N1100F059A MS East Bypass Valve Discrepancies
During 24.109.02
05/17/2021
29869
Received 4D91 Electric Governor Trouble During 24.109.02
09/25/2022
CR202436476
M&TE: Failed Calibration of MM215M (HP 34401A)
2/22/2024
CR202436477
F&TE: Failed Calibration of MM299M (Agilent U1253B
DMM)
2/22/2024
CR202436479
M&TE: Failed Calibration of PG896M (30 psig Crystal
Pressure Gauge)
2/22/2024
CR202436626
M&TE Backlog Identified
2/29/2024
CR202436769
Nuclear Oversight Area of Concern-M&TE Program
Appears at Risk for the Station and Could Adversely Impact
RF22
03/07/2024
CR202436828
Nuclear Oversight: Expired M&TE Equipment Left in the
Decontamination Room on TB2
03/11/2024
CR202437005
Gaps in M&TE Program Compliance Over Last 4 Years
Resulting in Presumed Lost Test Equipment and Other
Issues
03/19/2024
CR202437011
Nuclear Oversight - Numerous Pieces of M&TE Found
Sitting in I&C Hot Shop with RAM Tags From 2022
03/19/2024
CR202437148
Automatic RPS Scram on High RPV Pressure While
Attempting to Lower Generator MW to
MWe per 22.000.04
03/23/2024
CR202438279
E5150F007 - Non-Conservative Gear Ratio Discrepancy
Resulting in Negative Analytical Torque Margin
04/17/2024
CR202438319
E5150F007 Stem Nut-Wear Indicates Stem Nut Should Be
Replaced
04/16/2024
CR202438490
P4400F808-Non-Conservative Gear Ratio Discrepancy
Identified
04/17/2024
CR202438498
Potential MOV Gear Ratio Discrepancy EOC
Discovery-WO Request to Inspect P4400F802A * F802B
04/17/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202439109
E5150F007 Spring Pack Anomaly on AsLeft Thrust Testing
05/01/2024
6M7212044
Diagram Reactor Core Isolation Cooling System
BE
Drawings
6M7212045
Diagram Reactor Core Isolation Cooling System Barometric
Condenser
AT
Engineering
Changes
EDP 9890
Main Turbine Control System Mod F Valve Trip
06/01/1989
Post Scram Data
and Evaluation
Post Scram Data and Evaluation
03/24/2024
TSR100552
Technical Service Request: E5150F007 Valve Stroke Time
Increase
VMT11.6.16.1
GEC/New Parks Electro-Hydraulic Governor Book 1 of 3
C
VMT11.6.16.2
GEC/New Parks Electro-Hydraulic Governor Book 2 of 3
C
Miscellaneous
VMT11.6.16.3
GEC/New Parks Electro-Hydraulic Governor Book 3 of 3
C
24.109.02
Turbine Bypass Valve Operability Test
46.111.200
65030X-3148 Digital to Analog Converter Circuit Board
Calibration
47.110.01
Main Turbine Electro-Hydraulic Control System
Performance Test
ARP 4D91
Alarm Response Procedure 4D91 Electric Governor Trouble
MMA04
M&TE Program
Procedures
MMA05
Tool and M&TE Issue Control and Return
55766202
Change Division 1 Control Air Setpoints per DC 80149
04/27/2021
58403207
IST Required Perform MOV Diagnostic Testing
(G3352-F001)
05/09/2022
Work Orders
65708893
Install New Components for (EDP 80165 - Electrical)
2/20/2023
CR202334826
Tech Spec/TRM/ODCM Inadequate Change
Management - Licensed Operators Not Informed of
Changes
2/02/2023
CR202437054
Tech Spec/TRM/ODCM Inadequate Change
Management - Licensed Operators Not Informed of
Changes
03/21/2024
71152S
Corrective Action
Documents
CR202439929
CR202436854 Disposition Inadequate for Tracking TRM
Fire Protection Surveillance Requirements
05/30/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR202440178
NRC Identified: Unrestrained Mobile Scaffolds in RB
Basement NW Quad
06/10/2024
Corrective Action
Documents
Resulting from
Inspection
CR202440534
NRC Identified - All Drip Catches in Plant Not Tagged
06/26/2024
67.000.108
Catches and Containments
A
MLS08
Licenses, Plans, and Programs
Procedures
ODE8
Administrative Guidelines and Desk Instructions
24