IR 05000321/2025002
| ML25223A003 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/14/2025 |
| From: | Alan Blamey NRC/RGN-II/DORS/PB3 |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| IR 2025002 | |
| Download: ML25223A003 (1) | |
Text
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000321/2025002 AND 05000366/2025002
Dear Jamie M. Coleman:
On June 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch Nuclear Plant units 1 and 2. On August 13, 2025, the NRC inspectors discussed the results of this inspection with Kevin Carter, Regulatory Affairs Manager and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of the findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Edwin I. Hatch.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Edwin I. Hatch Nuclear Plant units 1 and 2.
August 14, 2025 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Alan J. Blamey, Chief Reactor Projects Branch 3 Division of Operating Reactor Safety Docket Nos. 05000321 and 05000366 License Nos. DPR-57 and NPF-5
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000321 and 05000366
License Numbers:
Report Numbers:
05000321/2025002 and 05000366/2025002
Enterprise Identifier:
I-2025-002-0028
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Edwin I. Hatch Nuclear Plant Units 1 and 2
Location:
Baxley, GA
Inspection Dates:
April 01, 2025 to June 30, 2025
Inspectors:
R. Easter, Resident Inspector
J. Lizardi-Barreto, Reactor Inspector
P. Niebaum, Senior Resident Inspector
J. Parent, Senior Resident Inspector
C. Scott, Senior Project Engineer
Approved By:
Alan J. Blamey, Chief
Reactor Projects Branch 3
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Edwin I. Hatch Nuclear Plant units 1 and 2 (Hatch), in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors.
Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Determine the Cause for a Significant Condition Adverse to Quality, and to Take Corrective Action to Preclude Repetition Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000366/2025002-01 Open/Closed
[P.1] -
Identification 71152A A self-revealed Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the licensees failure to determine the cause of the loss of unit 2 primary containment integrity identified on February 7, 2023, and take corrective actions to preclude repetition. This resulted in a repetitive loss of unit 2 primary containment integrity when torus purge supply valves, 2T48-309 and 2T48-324 exceeded the allowable leakage rate limit defined in Technical Specifications (TS) during local leak rate testing (LLRT) on February 22, 2025.
Failure to Perform Maintenance in Accordance with Work Instructions Appropriate to the Circumstance Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000366/2025002-02 Open/Closed
[H.1] -
Resources 71152A A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for work instructions that were inappropriate to ensure proper installation of the emergency overspeed switch (EOS) for safety-related emergency diesel generators (EDGs). As a result, the unit 2 A (2A) EDG was rendered inoperable since its EOS was installed on June 6, 2024, and when the improper installation was identified during a failed surveillance run on February 13, 2025.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000366/2024011-03 Seismic Qualification of Batteries with Cracked Cell Lids and Potential Part 21 for Nonconforming Epoxy Construction 71111.21M Closed URI 05000366/2023011-03 Potential Failure to Make a Part 21 Report 71152A Closed
LER 05000366/2025-001-00 Unit 2 - Primary Containment Inoperable Due To Exceeding Technical Specification Allowable Leakage Limits 71153 Closed LER 05000366/2025-002-00 Unit 2 - 2A Emergency Diesel Generator Inoperable Due to Inadvertent Overspeed Trip 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at or near rated thermal power (RTP). On April 5, 2025, power was lowered to approximately 70% RTP for turbine valve testing and was returned to full RTP on April 7, 2025. On May 31, 2025, power was lowered to approximately 70% RTP for turbine valve testing and was returned to full RTP on June 2, 2025. The unit remained at or near RTP for the remainder of the inspection reporting period.
Unit 2 began the inspection period at or near RTP. On May 24, 2025, power was lowered to approximately 70% RTP for turbine valve testing and was returned to full RTP on May 26, 2025.
On June 10, 2025, power was lowered to 92% RTP due to rising condensate temperatures after the loss of half of it's cooling towers. The unit was returned to RTP on nightshift of June 10, 2025. On June 27, 2025, power was lowered to approximately 35% RTP after the loss of the 'B' adjustable speed drive and trip of the 'B' recirculation pump. The unit remained at or near 35%
RTP for the remainder of the inspection reporting 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
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of hot weather for the following systems on June 10, 2025.
1. High voltage switchyard
2. Intake structure
3. Unit 1 and unit 2 turbine building cooling towers
4. Unit 1 and unit 2 reactor and radwaste building cooling towers
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) Unit 1 'C' emergency diesel generator (EDG) system on May 7, 2025
- (2) Unit 1 division 1 of residual heat removal (RHR) system on May 13 through 14, 2025
- (3) Unit 1 'B' EDG system on June 17, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 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) Diesel generator building on April 2, 2025
- (2) Intake structure on April 3, 2025
- (3) Unit 2 drywell chiller room, combustible materials without a transient combustible permit (Condition report [CR] 11170597), reviewed on May 8, 2025
- (4) Unit 1 turbine building 112' elevation on June 9, 2025
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on April 29, 2025.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated the readiness and performance of:
- (1) Work order (WO) SNC1397505, unit 1 'C' EDG heat exchanger inspections on June 16 through June 20, 2025
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the main control room during the following.
- Unit 2 power ascension operations on April 1, 2025
- Unit 2 'B' core spray valve operability on May 19, 2025
- Unit 1 scram discharge volume isolation valve functional test on May 20, 2025
- Unit 2 'B' residual heat removal service water pump in service testing on June 6, 2025
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated 25EX01 practice Emergency Management Program Evaluation drill on May 6, 2025.
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 remain capable of performing their intended function:
- (1) CR 11178143. Cognitive trend identified for reactor protection system power monitor output breakers reviewed June 9, 2025
- (2) CR 11185016. Main Steam analog transmitter trip system card found out of tolerance reviewed June 24, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 2 elevated risk with the 2E startup auxiliary transformer out of service for planned maintenance, on April 21, 2025
- (2) Unit 2 elevated risk and use of risk informed completion time (RICT) while a low condenser vacuum PCIS function was inoperable on April 28, 2025
- (3) Unit 1 elevated risk and use of RICT while the 'B' loop of the RHR system was out of service for maintenance from May 12 to May 27, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) CR 11162155. Unit 2 main steam line leak detection temperature element 2U61N112B reading high
- (2) CR 11171260. Unit 1 standby gas treatment damper 1T46-F001B exceeded its calculated allowable time limit
- (3) CR 11175488. Unit 1 drywell floor drain leakage rate
- (4) CR 11154299. Unit 2 loss of primary containment at penetration 2T23-X205
===71111.21M - Comprehensive Engineering Team Inspection Structures, Systems, and Components (SSCs) (IP section 03.01) (1 Partial)
(1)
(Partial)
The inspectors evaluated unresolved item (URI)05000366/2024011-03, "Seismic Qualification of Batteries with Cracked Cell Lids and Potential Part 21 for Nonconforming Epoxy Construction" (ADAMS Accession No. ML24354A088). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71111.21M. This URI is closed.
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)===
- (1) Procedures 34SV-E11-002-1, RHR valve operability, 34SO-E11-010-1, RHR system, and 34SV-E11-012-1, RHR shutdown cooling logic system functional test after unit 2
'B' train RHR system outage on May 27, 2025 (2)34SV-R43-006-1, EDG 1C fast-start test following a diesel system outage on June 21, 2025
Surveillance Testing (IP Section 03.01) (1 Samples)
(1)52SV-R43-001-0, unit 1 'C' EDG overspeed trip test on June 16, 2025
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1)34SV-T46-002-1, standby gas treatment system damper operability on April 18, 2025
- (2) 34SV-P41-001-1, unit 1 'B' loop of plant service water pump operability on June 24, 2025
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
- (1) The inspectors observed an emergency preparedness drill on May 6, 2025. The drill involved a carbon dioxide discharge in the EDG building, a loss of feed and reactor scram, a reactor water cleanup system line break in the drywell, a failure of a drywell penetration that results in a radiological release to the reactor building with increasing radiation levels that leads to a declaration of a General Emergency and issuance of protective action recommendations.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===
- (1) Unit 1 (April 1, 2024, through March 31, 2025)
- (2) Unit 2 (April 1, 2024, through March 31, 2025)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
- (1) Unit 1 (April 1, 2024, through March 31, 2025)
- (2) Unit 2 (April 1, 2024, through March 31, 2025)
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) CR 111600896. Unit 1 loss of B 24/48-volt (direct current) electrical cabinet (1R25S016) resulted in containment isolation actuations on March 14, 2025
- (2) Corrective action report (CAR) 867865. Unit 2 'A' EDG overspeed trip during testing on February 13, 2025
- (3) CAR 829588. Unit 2 local leak rate testing failure resulted in loss of primary containment on February 22, 2025
71152S - 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 packing leaks associated with safety related valves that might be indicative of a more significant safety issue.
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) to verify the accuracy and completeness of the information provided and the appropriateness of the corrective actions.
- (1) LER 05000366/2025-001-00, Unit 2 Primary Containment Inoperable Due To Exceeding Technical Specification Allowable Leakage Limits submitted April 7, 2025 (ADAMS Accession No. ML25107A275). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152A.
This LER is closed.
- (2) LER 05000366/2025-002-00, 2A Emergency Diesel Generator Inoperable Due to Inadvertent Overspeed Trip submitted May 23, 2025 (ADAMS Accession No. ML25143A260). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152A. This LER is closed.
INSPECTION RESULTS
Unresolved Item (Closed)
Seismic Qualification of Batteries with Cracked Cell Lids and Potential Part 21 for Nonconforming Epoxy Construction URI 05000366/2024011-03 71111.21M
Description:
The inspectors resolved an unresolved item (URI) concerning cracks in battery cell lids (i.e., covers) on the Hatch unit 2 trains A and B Class 1E station battery cells. The inspectors reviewed design and licensing basis documents, industry standards, and interviewed licensee staff in the review of this URI.
With respect to the identification of the degraded condition observed in the field, the inspectors found that the licensees failure to document abnormal conditions required by station procedures represented a performance deficiency and minor violation. This assessment is captured in the results section of this report as minor violation Failure to Document Deterioration on Battery Covers Due to Misleading Procedural Instructions.
With respect to the suitability of the repair materials, the inspectors found that the licensees failure to procure and install repair materials in accordance with quality-related processes represented a performance deficiency and minor violation. This assessment is captured in the results section of this report as minor violation Failure to Control Design and Quality Standards for Epoxy Repairs on Cracked Lids of Safety Related Batteries.
With respect to Title 10 of the Code of Federal Regulations (10 CFR) Part 21, Reporting of Defects and Noncompliance concerns, the inspectors found that the degraded conditions (i.e., cracked lids on the battery cells) did not represent a challenge to any license basis safety limits. Therefore, this assessment did not identify any associated findings or violations.
With respect to the seismic qualification concern, the inspectors did not identify that the degraded conditions observed adversely impacted the capability of the batteries to perform their safety function during a design basis event. Therefore, this assessment did not identify any associated findings or violations.
Corrective Action Reference(s): Condition reports (CRs) 11123451, 11138025, 11126278, 11197187, 11197204 Minor Violation: Failure to Document Deterioration on Battery Covers Due to Misleading Procedural Instructions 71111.21M Minor Violation: The inspectors identified a minor violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, for the licensees failure to document physical damage or abnormal conditions (i.e., cracks on battery cell lids) for multiple cells of the Unit 2 Class 1E station batteries, as required by battery surveillance procedures 52SV-R42-001-2, 52SV-R42-002-2, and of 52SV-R42-004-0. The licensee captured this issue in condition reports (CRs) 111234451, 11197204 and 11138025.
The licensees failure to document physical damage or abnormal conditions in accordance with quality related procedures was a performance deficiency.
Screening: The inspectors determined the performance deficiency was minor. The performance deficiency was not a precursor to a significant event, would not have led to a more significant safety concern if left uncorrected, and did not adversely impact the Mitigating Systems cornerstone objectives.
Enforcement:
This failure to comply with 10 CFR 50, Appendix B, Criteria V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Minor Violation: Failure to Control Design and Quality Standards for Epoxy Repairs on Cracked Lids of Safety Related Batteries 71111.21M Minor Violation: The inspectors identified a minor violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, for the licensees failure to procure and install materials in accordance with quality related processes for the repair of cracks in the lids of the Unit 2 Class 1E station batteries. Specifically, the licensee used an epoxy that was incompatible with acidic environments and did not review its suitability for use with lead-acid battery cells.
Additionally, there was no commercial grade qualification report for its application on Class 1E electrical components. The licensee captured this issue in condition reports (CRs)11138025, 11126278, 11197187 and technical evaluation (TE) 1165459.
The licensees failure to procure, evaluate, and use materials that were suitable for the repair of cracked lids on Class 1E station batteries was a performance deficiency.
Screening: The inspectors determined the performance deficiency was minor. The performance deficiency was not a precursor to a significant event, would not have led to a more significant safety concern if left uncorrected, and did not adversely impact the Mitigating Systems cornerstone objectives.
Enforcement:
This failure to comply with 10 CFR 50, Appendix B, Criteria III, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Failure to Determine the Cause for a Significant Condition Adverse to Quality, and to Take Corrective Action to Preclude Repetition Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000366/2025002-01 Open/Closed
[P.1] -
Identification 71152A A self-revealed Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the licensees failure to determine the cause of the loss of unit 2 primary containment integrity identified on February 7, 2023, and take corrective actions to preclude repetition. This resulted in a repetitive loss of unit 2 primary containment integrity when torus purge supply valves, 2T48-309 and 2T48-324 exceeded the allowable leakage rate limit defined in Technical Specifications (TS) during local leak rate testing (LLRT) on February 22, 2025.
Description:
On February 22, 2025, during refueling outage 2R28, Hatch unit 2 primary containment isolation valves (PCIV) 2T48-F309 and 2T48-F324 exceeded the administrative limit of 4,850 sccm during LLRT, with an as-found leak rate of 309,175 sccm. These 18-inch torus purge supply valves (F309 inboard and F324 outboard) individually surpassed the maximum TS allowable leakage (La), failing to maintain primary containment integrity through penetration 2T23-X205. The licensee determined that the operability of penetration 2T23X205 was lost sometime after the valves passed LLRT during refueling outage 2R27 in 2023 and sometime prior to the shutdown for 2R28 in 2025. Plant data showed that excessive torus purge supply valve leakage existed during the last operating cycle for longer than allowed by TS 3.6.1.1 Primary Containment.
The licensee submitted licensee event report (LER) 05000366/2025-001-00, Unit 2 Primary Containment Inoperable Due To Exceeding Technical Specification Allowable Leakage Limits (ADAMS Accession No. ML25107A275) on April 7, 2025, for this event. The inspectors reviewed the LER, and the licensees causal evaluation, as documented in corrective action report (CAR) 829588. The licensee determined that the root cause for the LLRT failures of F309 and F324, was that the valves did not seat properly due to excessive clearances between the T-Ring to disc interface. The excessive clearance, identified as a design issue, required a greater degree of T-Ring compression ring adjustment to achieve a proper seal. The increased mechanical stress resulted in relaxation of the T-Ring in the form of creep, resulting in separation between the T-Ring and disc.
On February 7, 2023, PCIVs 2T48-F319 and 2T48-F320 (same design valves as F309 and F324) failed their LLRT and exceeded the TS allowable leakage rate limit. This also represented a loss of unit 2 primary containment integrity. The 2023 causal evaluation for this issue, CAR 380222, determined that the root cause of the LLRT failure was the use of defective (i.e., rolled) T-Rings that were more susceptible to relaxation. The extent of condition evaluation completed in 2023 concluded that the root cause was also applicable to F309 because it also had the defective T-Ring. Valve F324 was determined to not have the defective 'T-Ring' installed. CAR 380222 determined that the LLRT failure and loss of unit 2 primary containment integrity was a significant condition adverse to quality (SCAQ). Section 2.0 Definitions of licensee procedure NMP-GM-002-001, Corrective Action Program Instructions, defines a SCAQ as a condition which, if uncorrected, could have a serious effect on nuclear safety or operability. In the case of a SCAQ, the measures shall assure that the cause of the condition is determined and a CAPR [corrective action to prevent recurrence]
implemented. The CAPR identified in 2023 included discontinuing the use of defective T-Rings and ensuring that Hatch had the correct design specifications for T-Rings per technical evaluation (TE) 1127350. These CAPRs were completed on May 25, 2023, but failed to prevent an LLRT failure for F309 and F324 on February 22, 2025, resulting in the loss of unit 2 primary containment integrity.
The inspectors determined that the licensee had multiple opportunities to identify the excessive clearance design issue during and prior to the 2023 root cause evaluation (CAR 380222) of the F319 and F320 valve failures. Specifically, the licensee conducted a causal analysis in 2017 (failures of valves F309, F319, and F320) and a root cause analysis in 2021 (failures of F319 and F320), both of which involved failure mechanisms related to the TRing-todisc interface. Although these evaluations identified contributing factors such as creep, cold flow, and compression set that adversely affected valve performance, the critical issue of excessive clearance was not recognized. Therefore, the inspectors concluded that the licensee had sufficient information to identify the clearance issue and implement effective corrective actions prior to the 2025 valve failures.
Additionally, an unresolved item (URI) associated with the failures of unit 2 primary containment isolation valves F319 and F320 was opened in IR 2023011 (ADAMS Accession No. ML23320A059) to determine if the licensee satisfied the 10 CFR Part 21 reporting requirement with the information provided in LER 05000366/2023-001-00 (ML23216A113)despite never conducting a substantial safety hazard evaluation for defective T-Rings. The LER stated that valves F319 and F320 exceeded the maximum TS allowable leakage (La)due to incorrectly manufactured T-Rings provided by the original equipment manufacturer.
As documented in this NCV, the licensees corrective actions to discontinue the use of defective T-Rings were ineffective in preventing the reoccurrence of a loss of unit 2 primary containment integrity. Also, the 2025 causal evaluation (CAR 829588) documented that a contributing cause to the loss of unit 2 primary containment integrity was that previously identified causes for T48 valve failures (in 2017, 2021, and 2023) masked the identification of the excessive clearance root cause. Based on this information, the inspectors could not determine if a violation exists or if defective T-Rings in F319 and F320 created a substantial safety hazard, as the failures may have been caused by excessive clearance in the T-Ring to disc interface identified in the 2025 root cause evaluation, following the failures of F309 and F324. URI 05000366/2023011-03 is closed.
Corrective Actions: Valve maintenance and a successful as-left LLRT was performed on each torus purge supply valve. Primary containment was restored to operable status. Additionally, anticipated T-Ring relaxation over the upcoming operating cycle was accommodated by tightening compression ring set screws on potentially impacted valves to maximize their closing torque.
Corrective Action References: Root cause report (CAR 829588).
Performance Assessment:
Performance Deficiency: The licensees failure to determine the cause for the loss of unit 2 primary containment integrity, a significant condition adverse to quality, and take corrective action to preclude a repetitive loss of primary containment integrity was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the containment isolation function of penetration 2T23X205 was not preserved during the previous operating cycle as evidenced by failed asfound LLRT of the torus purge supply valves in this penetration.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, Exhibit 3 C. Reactor Containment. Because this issue represented an open pathway in containment, the inspectors answered Yes to Question 1, which directed the inspectors to IMC 0609 Appendix H, Containment Integrity SDP. Screening the performance deficiency resulted in no impact to core damage frequency. Therefore, it was screened as a Type B Finding per Table 4.1 which list containment isolation valves, in large lines connecting the BWR Mark 1 containment airspace to the environment, as SSCs considered for large early release frequency (LERF) Implications. This screened to Table 7.1 which indicates to perform a Phase 2 assessment per Table 7.2. Because the as found leakage rate of 309,175 sccm represents approximately 1.13 La and therefore is not considered a LERF significant leakage rate from containment, the issue screens to Green.
Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to properly identify the root cause of PCIVs failing LLRT in 2023, and this led to LLRT failures of similar valves in 2025.
Enforcement:
Violation: 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires in part, that in the case of significant conditions adverse to quality, measures shall assure that the cause of the condition is determined, and corrective action taken to preclude repetition.
Hatch Nuclear Plant unit 2 Technical Specification (TS) Limiting Condition of Operation 3.6.1.1, Primary Containment, required primary containment shall be operable in Modes 1, 2 and 3. Also, it requires primary containment to be restored within one hour (Condition A.1),otherwise 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 /> (Condition B.1).
Contrary to the above, on May 25, 2023, the licensee failed to determine the cause of a significant condition adverse to quality (SCAQ) associated with the unit 2 primary containment system and take corrective actions to preclude repetition. Specifically, the licensee failed to determine the cause of the excessive leak rate in valves F319 and F320 after these exceeded the LLRT acceptance criteria on February 7, 2023. This failure resulted in the loss of unit 2 primary containment integrity and constituted an SCAQ as defined in procedure NMP-GM-002-001. As a result, the licensees corrective actions did not address the cause of the excessive leak rate and failed to preclude a subsequent loss of primary containment integrity due to the same failure mechanism, as demonstrated by the failure of valves F309 and F324 to meet the LLRT acceptance criteria on February 22, 2025.
Consequently, unit 2 primary containment was not operable in that leakage rates exceeded the maximum TS allowable primary containment leakage limits (La) for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while in mode 1 and the unit was not placed in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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: 05000366/2023011-03.
Failure to Perform Maintenance in Accordance with Work Instructions Appropriate to the Circumstance Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000366/2025002-02 Open/Close
[H.1] -
Resources 71152A A self-revealed Green finding and associated non-cited violation (NCV 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for work instructions that were inappropriate to ensure proper installation of the emergency overspeed switch (EOS) for safety-related emergency diesel generators (EDGs). As a result, the unit 2 A (2A) EDG was rendered inoperable since its EOS was installed on June 6, 2024, and when the improper installation was identified during a failed surveillance run on February 13, 2025.
Description:
On February 13, 2025, with unit 2 shut down and in Mode 5 for a planned refueling outage, the licensee conducted a test that involved running the 2A EDG.
Approximately 11 minutes into the test, the diesel tripped due to an inadvertent overspeed signal from the EOS, rather than an actual overspeed condition. On May 23, 2025, the licensee submitted licensee event report (LER) 05000366/2025-002-00, 2A Emergency Diesel Generator Inoperable Due to Inadvertent Overspeed Trip, (ADAMS Accession No. ML25143A260). The inspectors reviewed the LER, the licensees causal product, as documented in corrective action report (CAR) 821174, and discussed the issue with licensee staff.
The licensee determined (via failure analysis) that a loose retaining nut on the EOS plunger caused the plunger to shift away from the EDG overspeed pushrod during EDG operation.
During an overspeed condition, the pushrod retracts, and a spring in the EOS pushes the plunger outward. When the plunger reaches its trip setpoint (as it moves outward), it activates the overspeed relay and shuts down the diesel engine. The licensee determined that a fully loose retaining nut is not expected to cause an inadvertent overspeed trip if the EOS plunger is properly installed (i.e., halfway between the full engagement point with the engine overspeed pushrod and the overspeed trip actuation point). On February 16, 2011, the Fairbanks Morse Owners Group (FMOG) issued maintenance guidelines for replacement of the EOS, with the most recent revision (Revision 7) issued on July 21, 2022. Item S1 of this OE (titled Replacement of the Emergency Overspeed Switch) provides detailed instructions for replacing the EOS, including steps for proper installation and setting the switch plunger halfway between the full engagement and actuation points to ensure reliable switch activation and prevent false overspeed actuations, as experienced at other facilities. The licensee also identified that while the EOS plunger was confirmed to be engaged with the engine overspeed pushrod, the amount of overtravel was not recorded or confirmed to be in the halfway position, and if installed correctly, it would have increased margin for a false overspeed trip due to a fully loose retaining nut.
The 2A EDG EOS was last replaced using a preventive maintenance WO SNC1137787 on June 6, 2024. The inspectors reviewed the WO and noted that the EOS installation instructions relied on skill of the craft to set the plunger. However, details on how to properly set the plunger (i.e., halfway between pushrod full engagement and overspeed trip actuation point) were not included in the WO, and a review of the completed instructions provided no indication that this activity had been performed. The inspectors determined that the licensee did not incorporate this guidance into maintenance WO instructions for EOS installation on safety related EDGs. Based on the absence of proper EOS installation work order instructions (including documentation of as-left condition) and the subsequent trip of the 2A EDG with a loose EOS plunger retaining nut, the inspectors concluded that the EOS was likely not installed correctly on June 6, 2024.
Corrective Actions: The 2A EDG EOS was replaced on February 14, 2025, and tested satisfactorily. During the week of May 5, 2025, the licensee conducted an extent of condition inspection of all the EDGs that included re-installation with instructions incorporating the FMOG guidance and appropriate torque values.
Corrective Action References: CAR 821174, SNC2444993, SNC 2445000, SNC2445001, SNC245002, SNC2445003
Performance Assessment:
Performance Deficiency: The licensees failure to perform maintenance in accordance with work instructions that were appropriate to the circumstance for the safety related EDGs was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality 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, maintenance instructions for installation of the EOS for the safety related EDGs were inadequate. This contributed to the inoperability of the 2A EDG and an inadvertent overspeed trip of the 2A EDG on February 13, 2025.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the significance of the finding using Exhibit 2 of IMC 0609 Appendix A for the mitigating systems cornerstone and determined that a detailed risk evaluation was required because the failure of the 2A EDG EOS represented inoperability of one train of a multi-train system for greater than the Technical Specification allowed outage time.
A regional senior reactor analyst (SRA) conducted a detailed risk assessment of the degraded condition. The SRA used SAPHIRE 8 version 8.2.11 and the Hatch Units 1&2 SPAR model version 8.83 dated 9/7/2023, NRC Inspection Manual Chapters 0609 Appendix A, F, and H and the Risk Assessment Standardization Project Manual Volumes 1 and 2.
The SRA used the guidance in Risk Assessment Standardization Project Manual Volume 1 to establish the exposure time to be 110 days based on the run time of the diesel and the run time related loosening of the overspeed plunger retaining nut. The SRA made the following key assumptions. The failure of the EOS switch resulted in failure of the 2A EDG to run.
When an engine overspeed trip occurs several annunciators in the main control alert operators to the EDG overspeed condition. The Alarm response procedures direct the operators to recover the EDG in accordance with procedure 34AB-R43-001-2, DIESEL GENERATOR RECOVERY section 4.8.4 steps 8 and 9. These steps direct operators to recover the EDG by disconnecting the EOS and relying on operator action and the mechanical overspeed switch. Therefore, recovery was credited and had a high likelihood of success. FLEX Equipment and procedure were credited for recovery of Station Blackout events. The SRA adjusted the SPAR model to allow for recovery.
The licensee had made several plant modifications that were not modelled in the NRCs SPAR model but were part of the licensees risk models. As such, the SRA considered the licensees risk model to be the best available information for Internal Fire, Internal Flooding, Seismic, and Large Early Release calculations. The NRC SPAR model was considered the best available information for Internal Events and Tornado and High Winds sequences. The dominant accident sequence was a weather-related loss of offsite power event, with failures of the other EDGs, failure of the reactor core isolation cooling system and failure of operators to restore either onsite or offsite power in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Cumulative risk for internal and external events was less than 1 E-6 for change in core damage frequency and less than 1 E-7 for change in Large Early Release Frequency; therefore, this finding is characterized as very low safety significance (GREEN).
Cross-Cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. In this case, the licensee had access to maintenance guidance for the installation of EOS, and intended for maintenance technicians to implement it, however failed to provide an adequate procedure.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part, that activities affecting quality be performed in accordance with instructions, procedures, and drawings appropriate to the circumstance.
Hatch Technical Specification (TS) Limiting condition for operations (LCO) 3.0.1 requires, in part, that LCOs shall be met during the Modes in the TS Applicability section. TS LCO 3.8.1, AC Sources, required, in part, that two unit 2 diesel generators shall be operable in Modes 1, 2, and 3.
Contrary to the above on June 6, 2024, maintenance instruction affecting the 2A EDG EOS, were not appropriate to the circumstance. Work order SNC1137787 conducted on June 6, 2024, did not provide appropriate instructions to properly set the EOS plunger, which contributed to the failure of the 2A EDG during testing on February 13, 2025. Consequently, the 2A EDG was inoperable from June 6, 2024 until February 9, 2025 when unit 2 transitioned to Mode 4. This period of inoperability resulted in the 2A EDG being inoperable for longer than allowed by TS 3.8.1.B. Additionally, the 1B EDG (swing diesel) was also tagged out-of-service during this period for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and 49 minutes on June 21, 2024. The overlapping time of the inoperability of the 2A EDG and the 1B EDG resulted in two EDGs being inoperable for unit 2 longer than allowed by TS 3.8.1.F.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified that no proprietary information was retained or documented in this report.
- On July 22, 2025, the inspectors presented the results of the integrated inspection and review of URIs05000366/2023011-03 and 05000366/2024-001-03, to Matthew Busch, Site Vice President, and other members of the licensee staff.
- On August 13, 2025, the inspectors presented the results of the integrated inspection to Kevin Carter, Regulatory Affairs Manager.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
DI-OPS-56-0293
Hot Weather Operation
2.3
Procedures
NMP-GM-025
Seasonal Readiness Process
8.4
Procedures
NMP-OS-017
Severe Weather
5.0
Procedures
Residual Heat Removal System
46.9
Procedures
Diesel Generator Standby AC System
29.7
Corrective Action
Documents
CR 11169419
Combustible Materials in DWC Room without TCP
04/11/2025
Corrective Action
Documents
CR 11170597
MFIN Crew Learning - TCPs
04/16/2025
Corrective Action
Documents
CR 11174764
PA System not heard in areas of the plant during fire drill
04/30/2025
Corrective Action
Documents
CR 11174779
E Team Crew Learning - Unannounced Fire Drill
5/01/2025
Fire Plans
NMP-ES-035-
019-GL02-F06
Diesel Generator Building
1.0
Fire Plans
NMP-ES-035-
019-GL02-F07
Intake Structure
2.0
Fire Plans
NMP-ES-035-
019-GL02-F08
U1 Turbine Building El. 112
1.0
Fire Plans
NMP-ES-035-
019-GL02-F09
U1 Turbine Building El. 130
1.0
Fire Plans
NMP-ES-035-
019-GL02-F31
U2 Reactor Building El. 158/164
1.0
Fire Plans
NMP-TR-425-F14
Fire Drill Package
04/29/2025
Procedures
NMP-TR-425
Fire Drill Program
14.0
Corrective Action
Documents
CR 11188809
New motor amp high on 1C Jacket cooling pump
06/20/2025
Procedures
Molded Case Circuit Breaker Testing
2.32
Procedures
Diesel Generator Lube Oil Pumps Major Inspection/overhaul
06/18/2025
Procedures
Diesel, Alternator, and Accessories Inspection
06/20/2025
Procedures
NMP-MA-019
Bolting and Torque Guidelines
5.9
Work Orders
SNC1351811
1C EDG Jacket Coolant Pump Inspection
06/20/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Work Orders
SNC1397505
1C EDG system outage
06/17/2025
Work Orders
SNC1449905
Major Pump PM
06/17/2025
Procedures
25EX01 Practice
EMPE
Drill Package
05/06/2025
Procedures
Plant Startup
51.16
Procedures
Reactor Recirculation System
49.0
Procedures
Scram Discharge Volume Isolation Functional Test
05/20/2025
Procedures
RHR Service Water Pump Operability
06/06/2025
Procedures
Core Spray Valve Operability
05/19/2025
Procedures
Core Heat Balance
11.6
Procedures
NMP-RE-008-F01
Detailed Reactivity Management Plan
2.1
Corrective Action
Documents
CR 11185016
06/06/2025
Corrective Action
Documents
CR(s) 11178143,
11174091,
11176285,
11075324,
10986035,
10972196
Corrective Action
Documents
TE 1178494
Procedures
NMP-ES-027
11.1
Miscellaneous
Phoenix risk assessment system alignments
04/21/2025
Miscellaneous
2-RICT-2025-01
RICT report
04/28/2025
Procedures
NMP-GM-031
On-Line Configuration Risk Management Program
9.5
Procedures
NMP-GM-031-
003
Risk Management Actions for 10CFR50.65(a)(4) and the
Risk Informed Completion Time
9.2
Procedures
NMP-OS-010-002
Hatch Protected Equipment Logs
2.7
Corrective Action
Documents
CRs 11167623,
11167644,
11167652,
11167654
Corrective Action
Documents
11172871
IST Stroke Time Update in Operations Procedures
04/23/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Resulting from
Inspection
Drawings
H-11636
Unit 1 PI&D - H.P. Hydraulic Fluid Piping
Engineering
Evaluations
TE 1161934
Past Operability Review for CV#4 Failure
10/15/2024
Miscellaneous
34SV-T46-002-1 Standby Gas Treatment System Damper
Operability IST Reference Data Page
01/17/2023
Miscellaneous
ACMP - Elevated
U1 DWFD
Leakage Rate
Adverse Conditioning Monitoring Plan
05/14/2025
Miscellaneous
TE 1178231
Operational Decision Making Issue, 1-25-001
Procedures
Conditions, Required Actions, and Completion Times
8.13
Procedures
As found LLRT results for 2T48-F309, and F324
2/13/2025
Procedures
As found LLRT results for 2T48-F309
2/22/2025
Procedures
As left LLRT results for 2T48-F309, and F324
03/08/2025
Procedures
NMP-ES-013-
GL01
IST Positions
7.0
Corrective Action
Documents
CR 11123451
24 CETI - Lid Cracking Discovered on 2R42S001A
10/29/2024
Corrective Action
Documents
CR 11126278
24 CETI - NRC question concerning epoxy used on
battery lids
11/8/2024
Corrective Action
Documents
CR 11138025
NRC CETI Inspection URI - Seismic Qualification of
Batteries with Cracked Cell Lids and Potential Part 21 for
Nonconforming Epoxy Construction.
2/27/2024
Corrective Action
Documents
CR 11197204
NRC minor violation - failure to document battery lid
cracking
7/22/2025
Corrective Action
Documents
CR11197187
NRC minor violation - battery lid repair epoxy not evaluated
7/22/2025
Engineering
Evaluations
S-63279
Seismic Test Report for C&D Batteries With Cracked Cells
and Cracked Lids
Engineering
Evaluations
TE 1165459
24 CETI - NRC question concerning epoxy used on
battery lids
11/11/2024
Engineering
Evaluations
TR-00021
C&D Technologies' LCUN-33 (LCR-33) Cells & 2-Step
Battery Rack Assembly
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Guide for the Ventilation and Thermal Management of
Batteries for Stationary Applications
22
Miscellaneous
IEEE Recommended Practice for Maintenance, Testing, and
Replacement of Vented Lead-Acid Batteries for Stationary
Applications
2002
Miscellaneous
SNC1457940
Battery Inspection 2R42S001A
3/25/2024
Miscellaneous
SNC1713941
2SV-R42-001-2 Div 2 Battery Pilot Cell Cell Surveillance
8/16/2024
Miscellaneous
SNC1723211
2SV-R42-001-2 Div 1: Battery Pilot Cell Surveillance
8/27/2024
Procedures
Battery Pilot Cell Surveillance
19.5
Procedures
Battery/Individual Cell Surveillance
19.6
Procedures
Battery Inspection
9.12
Procedures
HE-S-07-001
Procurement Specification for 125 Volt Diesel Generator 1A
and 1C Batteries for E.I. Hatch Nuclear Plant - Unit 1
1.0
Procedures
NMP-ES-202
Functional Classifications of Components, Parts, and
Materials
Shipping Records
PO
SNC10292667
Purchase Order (PO) Number SNC10292667
8/16/2022
Corrective Action
Documents
111888997
Emergency Engine Shutdown Alarm
06/20/2025
Drawings
LFD-1-RPS-15
Logic Functional Diagram for Technical Specification 3.3.1.1-1, Item 9 Reactor Protection System Instrumentation
- Turbine Control Valve Fast Closure, Trip Oil Pressure -
Low
Miscellaneous
MIS-16-004
Units 1 & 2 Inservice Testing Program Fifth Inspection
Interval
8.0
Procedures
Pump and Valve Inservice Testing (IST)
16.1
Procedures
RHR Valve Operability
5/27/2025
Procedures
Residual Heat Removal System
05/27/2025
Procedures
Residual Heat Removal Shutdown Cooling LSFT
05/27/2025
Procedures
Plant Service Water Pump Operability
06/24/2025
Procedures
Diesel Generator 1C Monthly Test
06/19/2025
Procedures
Diesel, Alternator, and Accessories Inspection
06/16/2025
Procedures
Diesel, Alternator, and Accessories Inspection
06/16/2025
Work Orders
SNC1397505
06/16/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
NMP-EP-141
Event Classification
4.0
Procedures
NMP-EP-141-002
Hatch Emergency Action Levels and Basis
5.0
Procedures
NMP-EP-144
Protective Actions
71151
Procedures
Reactor Coolant Weekly (Sample Sheets for Unit 1 -
April 2024 through March 2025)
71151
Procedures
Reactor Coolant Weekly (Sample Sheets for Unit 2 - April
24 through March 2025)
Corrective Action
Documents
CRs 11600896,
11167602,
11167623, and
11167644
Corrective Action
Documents
CR 11159040
Cognitive Trend CR - 2E51F007 Packing Leaks
Procedures
MOV Electrical Backseating
6.4
Procedures
NMP-ES-007-008
Equipment Failure Trending
9.0
Procedures
NMP-GM-002
Corrective Action Program
19.1
Procedures
NMP-GM-051
Trending, Gap Identification, Analysis and Closure
1.0
Corrective Action
Documents
Cognitive Trend on Main Turbine Valve Fast Actuation
Failures
2/11/2024
Corrective Action
Documents
Appendix J Program - LLRT Penetration Failure Associated
with 2T23-X205 (2T48-F309, F324, & D006)
2/22/2025
Corrective Action
Documents
Documentation of a loss of safety function due to 1B/2A
EDG inop on 6/21/24
05/28/2025
Miscellaneous
LFD-1-RPS-14
Logic Functional Diagram for TS 3.3.1.1-1, Item 8 Reactor
Protection System Instrumentation - Turbine Stop Valve -
Closure
Miscellaneous
LFD-1-RPT-02
Logic Functional Diagram for TS 3.3.4.1.a.2 EOC-RPT, TCV
Fast Closure
0