IR 05000321/2025003
| ML25339A066 | |
| Person / Time | |
|---|---|
| Site: | Hatch, 07200036  |
| Issue date: | 12/12/2025 |
| From: | Alan Blamey NRC/RGN-II/DORS/PB3 |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| EA-NMSS-2023-0002, EAF-NMSS-2025-0213 IR 2025003, IR 2024001 | |
| Download: ML25339A066 (0) | |
Text
SUBJECT:
EDWIN I. HATCH - INTEGRATED INSPECTION REPORT 05000321/2025003 AND 05000366/2025003 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION 07200036/2024001 AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Jamie Coleman:
On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch. On December 2, 2025, the NRC inspectors discussed the results of this inspection with Matthew Busch, Site Vice President, and other members of your staff.
The results of this inspection are documented in the enclosed report.
Due to the temporary cessation of government operations, which commenced on October 01, 2025, the NRC began operating under its Office of Management and Budget-approved plan for operations during a lapse in appropriations. Consistent with that plan, the NRC operated at reduced staffing levels throughout the duration of the shutdown. However, the NRC continued to perform critical health and safety functions and make progress on other high-priority activities associated with the ADVANCE Act and Executive Order 14300. On November 13, 2025, following the passage of a continuing resolution, the NRC resumed normal operations.
However, due to the 43-day lapse in normal operations, the Office of Nuclear Reactor Regulation granted the Regional Offices an extension on the issuance of the calendar year 2025 inspection reports that should have been issued by November 13, 2025, to December 31, 2025. The NRC resumed the routine cycle of issuing inspection reports on November 13, 2025.
Two 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.
In addition, the NRC identified a violation of 10 CFR 72.48, paragraphs (c)(1), (c)(2), and (d)(1),
and provisions of 10 CFR 72.212 that resulted from a Certificate of Compliance (CoC) holders failure to comply with 10 CFR 72.48 for a CoC holder-generated design change to its multi-purpose canister (MPC) fuel basket, known as the continuous basket shim (CBS) variant, which altered the structural configuration from welded to bolted shims. However, an Interim Enforcement Policy (IEP) issued in August 2025 is applicable to this violation. Specifically, December 12, 2025 Enforcement Policy section 9.4, Enforcement Discretion for General Licensee Adoption of Certificate of Compliance Holder-Generated Modifications under 10 CFR Part 72.48, provides enforcement discretion to not issue an enforcement action for this violation. The licensee will be expected to comply with 10 CFR 72.212 provisions after the NRC dispositions the noncompliance for a CoC holder-generated change that affects the General Licensee.
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 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.
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, 05000366, and 07200036
License Numbers:
Report Numbers:
05000321/2025003, 05000366/2025003, and 07200036/2024001
Enterprise Identifier:
I-2025-003-0014 and I-2024-001-0125
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Edwin I. Hatch
Location:
Baxley, GA
Inspection Dates:
July 01, 2025 to September 30, 2025
Inspectors:
B. Bowker, Senior Reactor Inspector
R. Easter, Resident Inspector
D. Hardage, Senior Resident Inspector
J. Parent, Resident Inspector
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, 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
Uncorrected EHC Fluid Degradation Results in RPS Inoperability Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2025003-01 Open/Closed
[P.2] -
Evaluation 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality involving fluid degradation in the electro-hydraulic control system due to elevated ambient temperatures. This degradation impaired turbine control valve (TCV) pressure switches which rendered the associated reactor protection system (RPS) trip function inoperable, as required by Technical Specification (TS) Limiting Condition for Operation 3.3.1.1, Reactor Protection System Instrumentation.
Procedural Non-Compliance Results in a Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000366/2025003-02 Open/Closed
[H.12] - Avoid Complacency 71153 A Green, self-revealed finding and associated NCV of TS 5.4.1.a Written procedures was identified on September 13, 2025, for the licensees failure to implement procedural instructions during Unit 2 TCV testing. Specifically, operators did not perform a required step intended to prevent a RPS signal, resulting in the trip of both reactor recirculation pumps and a subsequent manual reactor scram.
Additional Tracking Items
Type Issue Number Title Report Section Status EDG EAF-NMSS-2025-0213 Interim Enforcement Policy associated with the Continuous Basket Shim 60855 Closed LER 05000321/2025-001-00 LER 2025-001-00 for Edwin I. Hatch Nuclear Plant, Unit 1 High Pressure Coolant Injection System Inoperable 71153 Closed
PLANT STATUS
Unit 1 operated at or near rated thermal power (RTP) at the beginning of the inspection period.
On July 10, 2025, the unit began a downpower in preparation for a planned maintenance outage (PMO) (H12025A) and was taken offline on July 11. Startup activities began on July 14, and the unit returned to RTP on July 19. On September 5, the unit experienced a recirculation runback signal, resulting in a power reduction to 79% RTP. On September 6, power was further reduced to approximately 68% RTP to support control rod pattern adjustments and turbine testing. The unit returned to RTP on September 9, 2025, and remained at that level for the remainder of the inspection period.
Unit 2 began the inspection period in single loop operation at 35% RTP. Power ascension began on July 2, 2025, following the recovery of the A reactor recirculation pump, and the unit reached RTP on July 5. On September 13, power was reduced to approximately 70% RTP to support control rod pattern adjustments and turbine testing. Later on September 13, the unit was manually scrammed from 70% RTP following the loss of both recirculation pumps during turbine testing (see NRC Event Notice 57928). Startup activities began on September 17, and the unit returned to RTP on September 22, 2025, where it remained for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 division 1 of plant service water (PSW) system on July 14, 2025.
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) Unit 1 residual heat removal (RHR) system July 11 through 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) Unit 1 steam chase on July 14, 2025.
- (2) Control building 112-foot elevation on August 18, 2025.
- (3) Unit 1 reactor building 185-foot elevation on August 19, 2025.
- (4) Unit 1 condensate storage tank enclosure on September 8, 2025.
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during a live-fire training on August 20, 2025.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the high-pressure coolant injection rooms of Units 1 and 2 on September 18 and 19, 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 Unit 1 main control room during plant startup following PMO H12025A on July 14, 2025
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated the "D" crew in the simulator during execution of scenario H-LT-AF-CPE-001122, titled "CPE Design - EHC, RFPT Oil Temp Cont, Cond Vac, ASD Cell, SDV, ATWS" on July 28, 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) Condition Report (CR) 11160784. Main control room door not functioning properly, reviewed on September 3, 2025.
- (2) CR 11195544. Maintenance Rule (i.e., 10 CFR 50.65) evaluation for Unit 1 turbine control valve (TCV) failures, reviewed on September 11, 2025.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 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 1 outage risk assessment during PMO H12025A on July 11, 2025.
- (2) Unit 2 elevated risk due to A core spray (CS) system outage, on July 8, 2025.
- (3) Unit 1 and 2 elevated risk due to 2C emergency diesel generator (EDG) system outage during the week of August 10-16, 2025.
- (4) Unit 1 elevated risk due to 1C startup auxiliary transformer being out of service for bushing repair from September 9-10, 2025.
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) CR 11167602. Unit 1 TCV no.1 failure to initiate fast-closure trip and recirculation pump trip (System A) trip during surveillance, reviewed on August 14, 2025.
- (2) Technical evaluation (TE) 1161934. Past operability review for Units 1 and 2 TCV surveillance failures, reviewed on August 19, 2025.
- (3) CR 11192093. Unit 2 stuck control rod 30-11, reviewed on August 21, 2025.
- (4) TE 1180235. Past operability review for a fuel leak on the 2A EDG, reviewed on August 26, 2025.
- (5) TE 1175694. Past operability review of the Unit 2 'B' loop RHR system after a breaker trip, reviewed on September 16, 2025.
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Design Change Package SNC544222. Replacement of EDG 1C battery chargers with new chargers having the same electrical ratings.
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated Unit 1 H12025A activities including the following replacement and repair activities from July 11, 2025, to July 14, 2025.
- PSW leak repairs
- Main steam isolation valve (MSIV) repairs
- (2) The inspectors evaluated Unit 2 forced outage activities following a manual reactor scram from September 13 to September 18, 2025.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (6 Samples)
- (1) Work Order (WO) SNC2693294. Functional test of Unit 1 "C" MSIV following maintenance on July 14, 2025.
- (2) SNC615863, and SNC2599398. Repair and functional test of Unit 1 drywell chillers, reviewed on July 16, 2025.
- (3) SNC2360002. Functional tests of the 1A EDG battery charger, reviewed on July 25, 2025.
- (4) SNC1037020. Functional test of the main control room ventilation system crosstie valve following maintenance, reviewed on July 28, 2025.
- (5) SNC1957950. Functional test of the Unit 1 "B" train CS system following maintenance, reviewed on August 28, 2025.
- (6) SNC2789082. Functional test of the Unit 2 "G" intermediate range monitor following maintenance, reviewed on September 15, 2025.
Surveillance Testing (IP Section 03.01) (4 Samples)
- (1) Procedure 34SV-R43-002-2. EDG 1B monthly test run from Unit 2 on July 18, 2025.
(2)34SV-C71-005-1. TCV fast-closure instrument functional test on April 5-6, 2025.
(3)34SV-E11-004-1. Unit 1 division 1 residual heat removal service water pump test on July 30, 2025.
(4)34SV-E21-002-2. Unit 2 B train of CS system valve test on August 19, 2025.
Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)
- (1) Monitoring of elevated unidentified leakage in Unit 1 that led to PMO H12025A.
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) WO SNC1726122. Functional test and inspection of the FLEX Mode 5 core cooling pump, reviewed on September 16, 2025.
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance during a continuing training dynamic simulator scenario conducted on July 7, 2025. The scenario involved a sequence of events including a lowering condenser vacuum, loss of the control rod drive system, an adjustable speed drive cell bypass, an anticipated transient without scram (i.e., ATWS), and injection of boron. The scenario resulted in an Alert emergency classification, including notification to the State of Georgia and surrounding counties.
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 1 (July 1, 2024, through June 30, 2025)
- (2) Unit 2 (July 1, 2024, through June 30, 2025)
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
- (1) Unit 1 (July 1, 2024, through June 30, 2025)
- (2) Unit 2 (July 1, 2024, through June 30, 2025)
MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)
- (1) Unit 1 (July 1, 2024, through June 30, 2025)
- (2) Unit 2 (July 1, 2024, through June 30, 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) CRs 11167602, 11167623, and 11167644. Unit 1 TCVs No. 1, 2, and 4 failed surveillance testing conducted on April 5 and 6, 2025.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated the failure of three out of four TCV pressure switches to actuate the associated reactor protection system trip during surveillance testing on Unit 1 due to degraded Electro-Hydraulic Control (EHC) fluid, and the licensees response on April 6, 2024. Per IMC 0309, Reactive Inspection Decision Basis for Power Reactors, the NRC considered whether this inspection sample should be the subject of a reactive inspection and determined that a reactive inspection was not necessary, as documented in NRC's Agencywide Documents Access and Management System (ADAMS) under Accession No. ML25106A142. The inspection conclusions associated with this surveillance failure are documented in this report under Inspection Results Section 71152A.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000321/2025-001-00, Edwin I. Hatch Nuclear Plant, Unit 1 High Pressure Coolant Injection System Inoperable (ML25210A397). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct, and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is Closed.
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees response and personnel performance during an uncomplicated reactor scram of Unit 2 on September 13, 2025, which was caused by a human performance-related procedural error.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60855 - Operation of an Independent Spent Fuel Storage Installation 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) 2690, Inspection Program for Storage of Spent Reactor Fuel and Reactor-Related Greater-than-Class C Waste at Independent Spent Fuel Storage Installations (ISFSI) and for 10 CFR Part 71 Transportation Packagings."
Operation of an Independent Spent Fuel Storage Installation===
- (1) The inspector conducted a periodic in-office follow-up that focused on the review of the licensees implementation of the 10 CFR 72.48 process and associated corrective actions related to ISFSI activities. The review included:
- 72.48 Evaluations and Screenings: Reviewed the licensees 72.48 process and associated evaluation associated with the adoption of the continues basket shim (CBS) variant.
- Corrective Action Program: Reviewed condition reports related to the design change of the CBS basket variant.
INSPECTION RESULTS
Enforcement Discretion Enforcement Action EAF-NMSS-2025-0213: Interim Enforcement Policy associated with the Continuous Basket Shim 60855
Description:
Holtec International (also referred to as the Certificate of Compliance (CoC)holder) implemented a design change to its multi-purpose canister (MPC) fuel basket, known as the continuous basket shim (CBS) variant, which altered the structural configuration from welded to bolted shims. This change resulted in a departure from the method of evaluation (MOE) described in the final safety analysis report (FSAR) used to establish the design basis for tip-over events. Holtec did not fully evaluate the cumulative impact of the MOE changes or apply them consistently within the licensing basis. As a result, the NRC issued three Severity Level IV violations to Holtec for non-compliance with 10 CFR 72.48 requirements (see NRC Inspection Reports 07201014/2022-201, Holtec International (ADAMS Accession No.
ML23145A175) and 07201014/2022-201, Holtec International, Inc. - Notice of Violation (ML24016A190).
When the licensee (also referred to as a General Licensee) chooses to adopt a change the CoC holder made pursuant to a CoC holder's change authority under 10 CFR 72.48 (referred to herein as a CoC holder-generated change), the licensee must perform a separate review using the requirements of 10 CFR 72.48(c). Accordingly, when the licensee chooses to adopt a CoC holder-generated change, and that change results in a non-conforming cask, there is a violation of 10 CFR 72.48 and certain provisions of 10 CFR 72.212 by the licensee, in addition to a CoC holder violation of 10 CFR 72.48.
In support of the 2023 loading campaign, the licensee adopted Holtecs generic design change, as documented in the Hatch 10 CFR 72.48 Screening/Evaluation LDCR 2023-009 Version 1.0 Hatch 10 CFR 72.212 Report - Revision for UFLO 24, and subsequently loaded casks using the CBS basket design. Because the CoC holder-generated change was found to be noncompliant by the NRC, the loaded casks at Hatch were also rendered non-conforming.
Corrective Actions: The licensee entered this into their corrective action program with actions to restore compliance with the 10 CFR 72.212 provisions that require each cask to conform to the terms, conditions, and specifications of a CoC or an amended CoC listed in 10 CFR 72.214.
Corrective Action References: CR 11045817, TE 1147113, and TE 1151550
Enforcement:
Significance/Severity: The licensees failure to request that the CoC Holder obtain an amendment prior to implementing the change was determined to be of Severity Level IV significance based on the guidance in section 1.2.6.D of the NRC's Enforcement Manual. The severity of the violation was determined based on its very low safety significance, as documented in NRC memorandum titled Safety Determination of a Potential Structural Failure of the Fuel Basket During Accident Conditions for the HI-STORM 100 and HI-STORM Flood/Wind Dry Cask Storage Systems, (ML24018A085) and its similarity with violation example 6.1.d.2 in the NRCs Enforcement Policy.
Violation: Title 10 CFR 72.48 (c)(1) requires, in part, that licensee or certificate holder may make changes in the facility or spent fuel storage cask design as described in the FSAR (as updated), without obtaining:
- (ii) CoC amendment submitted by the certificate holder pursuant to § 72.244 if:
- (c) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this section.
Title 10 CFR 72.48(c)(2) requires, in part, that a general licensee shall request that the certificate holder obtain a CoC amendment, prior to implementing a proposed change, if the change would: (viii) Result in a departure from an MOE described in the FSAR used in establishing the design bases or in the safety analyses.
Title 10 CFR 72.48(d)(1) requires, in part, that the licensee shall have a written evaluation which provides the bases for the determination that the change does not require a CoC amendment pursuant to 72.48(c)(2).
Title 10 CFR 72.212(b)(3) requires, in part, a general licensee must ensure that each cask used by the general licensee conforms to the terms, conditions, and specifications of a CoC or an amended CoC listed in 72.214.
Contrary to the above, since the 2023 loading campaign, the licensee failed to:
- (1) request Holtec, the certificate holder, obtain a CoC amendment for a change to the CBS cask design that resulted in a departure from an MOE described in the FSAR;
- (2) have a written evaluation providing the bases for the determination that the adopted change did not require a CoC amendment; and
- (3) ensure that the affected casks conformed to the terms, conditions, and specifications of the applicable CoC.
Specifically, Hatch's 10 CFR 72.48 Screening/Evaluation LDCR 2023-009 Version 1.0 Hatch 10 CFR 72.212 Report - Revision for UFLO 24, failed to identify that the CBS variant design change resulted in a departure from a method of evaluation described in the FSAR used in establishing the design bases, failed to request the certificate holder obtain a CoC amendment pursuant to 10 CFR 72.244, and failed to ensure each cask conformed to the terms conditions, and specifications of a CoC or an amended CoC listed in 72.214, prior to using the CBS variant design.
Basis for Discretion: Section 9.4 of the Enforcement Policy, titled "Enforcement Discretion for General Licensee Adoption of Certificate of Compliance Holder-Generated Changes under 10 CFR 72.48" (ML25224A097), states that NRC will exercise enforcement discretion and not issue an enforcement action to a GL, for a non-compliance with the requirements of paragraphs (c)(1) and
- (2) and (d)(1) of 10 CFR 72.48 and with provisions of 10 CFR 72.212 that require GLs to ensure use of casks that conform to the terms, conditions and specifications of a CoC listed in 10 CFR 72.214, when the non-compliance results from a CoC holders failure to comply with 10 CFR 72.48 for a CoC holder-generated change. In support of the 2023 loading campaign, the licensee adopted a generic CoC holder design change (the CBS basket variant) and subsequently loaded the casks. On January 30, 2024, the NRC issued a notice of violation to the CoC holder, identifying the non-compliance, for the generic design change associated with the CBS basket variant (ML24016A190). As a result, the licensee became noncompliant due to the CoC holders failure to comply with 10 CFR 72.48 for the CoC holder-generated change. Since this violation meets the criteria of Section 9.4 of the policy, the NRC is exercising enforcement discretion by not issuing an enforcement action for this violation.
Uncorrected EHC Fluid Degradation Results in RPS Inoperability Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2025003-01 Open/Closed
[P.2] -
Evaluation 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality involving fluid degradation in the electro-hydraulic control (EHC) system due to elevated ambient temperatures. This degradation impaired turbine control valve (TCV) pressure switches which rendered the associated reactor protection system (RPS) trip function inoperable, as required by Technical Specification (TS)
Limiting Condition for Operation (LCO) 3.3.1.1, Reactor Protection System Instrumentation.
Description:
The turbine EHC system provides high pressure fluid (EHC fluid) to various valves in the reactor pressure/turbine control system. On April 5 and 6, 2025, during turbine valve testing in accordance with licensee procedure 34SV-C71-005-1, Turbine Control Valve Fast Closure Instrument Functional Test, TCVs #1, #2 and #4 closed as expected; however, the associated RPS pressure switches failed to actuate due to sludge buildup in the TCV EHC system. The sludge was produced by thermal degradation of the EHC fluid. The fluid remains chemically stable below 140F; however, ambient temperatures near the TCVs exceed this limit when the plant operates at or near rated thermal power. The fast-acting solenoid (FAS) valves are particularly susceptible to sludge deposition because only a very small amount of fluid flows past the tight clearances on the FAS spool. Degraded fluid (i.e.,
sludge) cannot pass through these clearances and therefore accumulates within the valve.
This buildup impaired the FAS valves ability to properly port fluid from the RPS pressure switch sensing lines, preventing adequate pressure decay during testing and resulting in failure of the pressure switches to actuate, even though the TCVs mechanically closed as expected. The failure of the RPS pressure switches to actuate would have prevented initiation of a reactor scram on TCV fast-closure, which is the primary protection against pressure and power transients following a generator load rejection. While at power, the only available method to temporarily clear these deposits is to cycle the FAS valves.
The licensee has experienced similar failures since 2011. Despite implementing several corrective actions to improve EHC fluid quality (e.g., fluid replacement (2011), dry air blankets (2015), electrostatic filters and monthly sampling (2016)), failures persisted. After chemical cleaning in 2022, quarterly testing was successful for 18 months. However, when the testing interval was extended to six months in late 2023, failures resumed. Between December 1, 2023, and September 2024, three trend condition reports were generated, and a recommendation to return to quarterly testing was not approved.
Per procedure NMP-GM-002-001, Corrective Action Program Instructions, a condition that has the potential risk of inhibiting, or has inhibited, a non-safety-related SSC from performing a function necessary for a safety-related function is a condition adverse to quality (CAQ). The elevated temperatures near the TCVs inhibited the EHC systems ability to support the RPS trip function; moreover, the extent of sludge buildup in the system was known to be time and temperature dependent. However, corrective actions were not taken to ensure the affected equipment remained operable until the next surveillance, and the licensee continued to rely on temporary measures (manual cycling) rather than addressing the underlying cause.
The surveillance testing on April 5-6, 2025, identified inoperable RPS TCV pressure switch channels for TCVs 1, 2, and 4. Based on the last successful test on September 8, 2024, and the licensees history of successful quarterly testing following chemical cleaning, the inspectors concluded the channels likely remained operable for at least one quarter. Since failures recurred only after extending the test interval to six months, the inspectors selected February 6, 2025, the midpoint between December 8, 2024 (one quarter after the last successful test), and April 5, 2025, as a reasonable estimate of when inoperability began.
Corrective Actions: The licensee removed degraded EHC fluid from the system by manually cycling the FAS valves and successfully completed TCV surveillance testing. Interim actions include resuming quarterly main turbine valve testing. Also, the licensee will consider design changes to lower condenser bay temperatures and/or improve EHC system stagnant/dead leg sections.
Corrective Action References: Condition reports (CRs) 11167602, 11167623, 11167644, 1122063 and Corrective Action Report (CAR) 750723.
Performance Assessment:
Performance Deficiency: The failure to identify and correct a CAQ associated with elevated ambient temperatures in the vicinity of the TCV EHC system and pressure switches as required by 10 CFR Part 50, Appendix B, Criterion XVI, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, three of four safety-related TCV pressure switches that provide signals to the RPS scram actuation logic were rendered inoperable.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
specifically screening it through Exhibit 2, Section C, Reactor Protection System. Because multiple channels of the TCV RPS trip function were affected (i.e., inoperable) and the licensees safety analysis did not credit a redundant RPS trip to mitigate the most limiting overpressure event (i.e., a load reject with turbine bypass valves inoperable), the issue required a Detailed Risk Evaluation (DRE). However, since RPS continues to fulfill its Probabilistic Risk Assessment (PRA) function, the issue only impacts thermal limits.
Currently, PRA models do not evaluate thermal limit exceedances that do not challenge the RPS function (i.e., do not result in an ATWS scenario). Therefore, the finding was further screened using Exhibit 3, Section A, Fuel Cladding Integrity, Question A2 of IMC 0609, Appendix M, which is the most relevant, and directed a DRE using qualitative criteria.
A regional Senior Reactor Analyst (SRA) performed the DRE in accordance with IMC 0609 Appendix M. The SRA determined that there was no calculated increase in Core Damage Frequency (CDF) or Large Early Release Frequency (LERF) since the PRA function of both the turbine bypass valves and RPS system would still be met. The SRA considered the redundant RPS trips, such as the turbine stop valve trip, the reactor high pressure trip, and the reactor power range high flux trip. These trips were not affected by the performance deficiency and would still initiate a reactor trip if needed. The turbine stop valve trip is credited in the licensee's safety analysis for a turbine trip when the turbine bypass valves are unavailable. Additionally, the licensee applied a Minimum Critical Power Ratio penalty, as required by the Core Operating Limits Report, due to the declared inoperability of the turbine bypass valve fast-open feature, providing additional safety margin. The SRA also noted that the End-of-Cycle Recirculation Pump Trip (EOC-RPT) function, while not credited in the safety analysis, was functional during the exposure period. These factors made it very unlikely that thermal limits would have been exceeded. Therefore, the finding was determined to be of very low safety significance (i.e., Green).
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. In this case, the licensee had information that area temperatures in the vicinity of the Unit 1 TCVs and safety-related RPS pressure switches were a condition affecting the operation of inputs to RPS but failed to adequately assess the impacts of EHC fluid degradation and potential for sludge buildup in the EHC system.
Enforcement:
Violation: 10 CFR 50 Appendix B Criterion XVI Corrective Action, states, in part, measures shall be established to assure that CAQs, such as nonconformances are promptly identified and corrected.
Hatch Unit 1 TS LCO 3.3.1.1, Reactor Protection System Instrumentation, requires that the RPS instrumentation for each function in Table 3.3.1.1-1 shall be OPERABLE. Table 3.3.1.1 For Function 9, Turbine Control Valve Fast Closure, Trip Oil Pressure - Low the TS requires two channels per trip system. Condition A requires that, with one or more required channels inoperable, the affected channel be placed in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Condition D requires that, if the completion time of Condition A is not met, the licensee must enter Condition E, which requires the reduction of thermal power to less than 27.6 percent of rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Contrary to the above, since December 1, 2023, the licensee failed to identify and correct a CAQ associated with elevated ambient temperatures in the vicinity of the TCV EHC system and pressure switches. In addition, from February 6, 2025, to April 6, 2025, three Turbine Control Valve Fast Closure, Trip Oil Pressure - Low channels were inoperable. The inoperable channels were not placed in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and the units thermal power was not reduced to less than 27.6 percent within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, as required by TS Conditions A, D, and E. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low channels were inoperable due to degraded EHC fluid from February 6, 2025, and remained inoperable until April 6, 2025, when the TCV fast-acting solenoids were manually actuated to clear the accumulated degraded fluid.
Enforcement Action: This violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Procedural Non-Compliance Results in a Reactor Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000366/2025003-02 Open/Closed
[H.12] - Avoid Complacency 71153 A Green, self-revealed finding and associated NCV of TS 5.4.1.a Written procedures was identified on September 13, 2025, for the licensees failure to implement procedural instructions during Unit 2 TCV testing. Specifically, operators did not perform a required step intended to prevent a RPS signal, resulting in the trip of both reactor recirculation pumps and a subsequent manual reactor scram.
Description:
On September 13, 2025, the licensee reduced Unit 2 power to 70 percent rated thermal power to perform a control rod sequence exchange, control rod scram time testing, and turbine valve testing. During the performance procedure 34SV-C71-005-2, Turbine Control Valve Fast Closure Instrument Functional Test, operators were testing the final TCV
(#4) when both reactor recirculation pumps tripped. Procedure step 4.9.4 requires operators to place switch 2C71-S12A,RPT SYS A OUT OF SERVICE, in the RPT SYS A INOP position to block the trip signal to the reactor recirculation pumps during the TCV #4 test (step 4.9.15). Operators performing the procedure inadvertently failed to implement step 4.9.4, allowing the trip logic to remain active. When TCV #4 was tested, a valid trip signal was generated, resulting in the automatic trip of both reactor recirculation pumps. In response to the total loss of recirculation flow, operators initiated a manual reactor scram. All control rods fully inserted and the reactor was safely shut down. The licensee reported this event to the NRC under Event Notification 57928.
Corrective Actions: Post-event reviews identified inconsistencies in place-keeping practices with respect to concurrent verification, independent verification, and section completion documentation. The licensee conducted operations department standdowns and increased supervisory oversight with an emphasis on human performance tool usage. Additional training is being provided to all operations department personnel focusing on human performance tool usage.
Corrective Action References: CR 11211383 and CAR 893796.
Performance Assessment:
Performance Deficiency: The failure to block the RPS trip signal to the Unit 2 reactor recirculation pumps by placing switch 2C71-S12A in the RPT SYS A INOP position, as required by step 4.9.4 of turbine valve test procedure 34SV-C71-005-2, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance 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 performance deficiency led to an event where both Unit 2 reactor recirculation pumps tripped, requiring a manual reactor scram due to the loss of both reactor recirculation pumps, and upset plant stability.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 1, Initiating Events Screening Questions. the inspectors determined that although the finding did result in a reactor trip, it did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Therefore, the finding screened as very low safety significant (i.e., Green).
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. In this instance, the operations crew did not effectively use human performance error reduction tools, including place-keeping, to ensure steps used to block the reactor recirculation pump trips, a critical step of a TS-required surveillance procedure, was performed during turbine valve testing.
Enforcement:
Violation: Hatch Unit 2 TS 5.4.1.a Written procedures, states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, February 1978.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Paragraph 8.b, Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Procedures. and Calibrations, requires, in part, that implementing procedures are required for each surveillance test listed in the TS.
Licensee procedure 34SV-C71-005-2 Turbine Control Valve Fast Closure Instrument Functional Test, is a TS-required surveillance. Step 4.9.4 directs switch 2C71-S12A, RPT SYS A OUT OF SERVICE, be placed in the RPT SYS A INOP position.
Contrary to this requirement, on September 13, 2025, the licensee failed to implement step 4.9.4 of procedure 34SV-C71-005-2. The failure to implement the procedure resulted in the trip of both Unit 2 reactor recirculation pumps and the subsequent manual reactor scram.
Enforcement Action: This violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 2, 2025, the inspectors presented the integrated inspection results to Matthew Busch, Site Vice President, and other members of the licensee staff.
- On September 30, 2025, the inspectors presented the ISFSI inspection results to Clay Channell, Fleet Dry Cask Program Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Residual Heat Removal System
46.9
Procedures
Plant Service Water System
36.24
Corrective Action
Documents
CR 11103201
NRC Resident Concern - Fire Hose Testing Stickers
08/19/2024
CR 11203723
NRC Resident Concern. Fire Extinguisher resting against
the HCVS air piping
08/14/2025
Corrective Action
Documents
Resulting from
Inspection
CR 11204803
Procedure Enhancement / Update 52SV-FPX-023-0 Fire
Hose Hydrostatic Testing
08/19/2025
NMP-ES-035-
019-GL02-F01
Control Building El. 112
1.0
NMP-ES-035-
019-GL02-F13
U1 Reactor Building El. 130
1.0
NMP-ES-035-
019-GL02-F15
U1 Reactor Building El. 185
1.0
Fire Plans
NMP-ES-035-
019-GL02-F52
Unit 1 Condensate Storage Tank
2.0
S-FP-113
Structural Fire Control Annual Fire Training
08/21/2025
Miscellaneous
S-FP-PE-DLA001
Fire Brigade Practical Exercises
08/21/2025
Fire Hose Hydrostatic Testing
1.1
Procedures
NMP-TR-426
Fire Training Program
10.0
Work Orders
SNC1009081
2SV-FPX-023-0 Non-High Rad Areas Fire Hose Station -
Unit One Hydrostatic Testing
07/24/2023
HNP-1-FSAR-6
Unit 1 FSAR Section 6.4.1
Miscellaneous
HNP-2-FSAR-9
Unit 2 FSAR Section 9.3.3.2.2
Miscellaneous
Reactivity Plan - Sequence Identifier: A1SU
07/14/2025
Plant Startup
46.10
H-LT-AF-CPE-
00112
CPE Design - EHC, RFPT Oil Temp Cont, Cond Vac, ASD
1.3
NMP-AD-006
Infrequently Performed Tests and Evolutions
Version 13.3
NMP-RE-008-F01
Detailed Reactivity Management Plan - Hatch 1 Cycle 32
MOC32 July Startup
07/14/2025
Procedures
NMP-TR-405
Simulator Exercise Guide and Evaluation
2.1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR 11146242
Main Control Room Door Not Latching
01/28/2025
CR 11156466
MCR East entrance door needs a screw in the card reader
locking mechanism
03/01/2025
CR 11160784
Main Control Room Door
03/14/2025
Corrective Action
Documents
TE 1175528
NRC Resident Concern - MCR Door
04/01/2025
CR 11165812
NRC Resident Concern - Is the main control room door
included in the scope of the maintenance rule program
03/31/2025
Corrective Action
Documents
Resulting from
Inspection
CR 11195544
Maintenance Rule Evaluation
07/15/2025
TE 1182480
MRE Unit 1 Turbine Control Valve #1
09/04/2025
TE 1182483
Maintenance Rule Evaluation
09/03/2025
Engineering
Evaluations
TE 1185309
MRE Unit 1 Turbine Control Valve #2
09/04/2025
Procedures
Control Room Habitability Program
2.0
Work Orders
SNC2377478
Main Control Room Door
03/15/2025
Calculations
Unit 1 Phoenix
model
Unit 1 Phoenix model
09/09/2025
Miscellaneous
HNP.H12025A
H12025A Outage Critical Path
07/11/2025
Outage Safety Assessment
Version 5.6
NMP-DP-001
Operational Risk Awareness
Version 27.0
NMP-GM-031
On-Line Configuration Risk Management Program
Version 9.5
NMP-OM-002
Shutdown Risk Management
Version 6.1
Procedures
NMP-OS-010-002
Hatch Protected Equipment Logs
2.8
Work Orders
WO SNC1116643
2A CS System Outage UT Voids Testing
07/08/2025
CR 11150993
2E11F048B will not throttle open or closed
2/12/2025
CR 11155310
RHR B Loop HX Bypass not functioning correctly
2/25/2025
CR 11164533
2E11F048B Breaker Trip
03/26/2025
CR 11167602
CV#1 on Unit 1 failed to cause fast closure trip, and RPT
System A trip
4/5/2025
CR 11187064
2A EDG Fuel Oil Header Line Leak
06/13/2025
CR 11192093
Stuck Control Rod on Unit 2
07/02/2025
Corrective Action
Documents
TE 1161934
Past Operability Review for CV#4 failure
09/18/2024
TE 1175694
POR: 2E11F048B Breaker Trip
05/06/2025
Engineering
Evaluations
TE 1182188
MRE - Stuck Control Rod on U2
07/16/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Operability
Evaluations
TE 1180235
EDG engineer to perform past operability review (POR)
08/03/2025
Inability To Move Control Rod
2.4
Procedures
Control Rod Drive Hydraulic System
34.4
Miscellaneous
HE-S-10-001
Specification for Battery Chargers
6.0
Work Orders
SNC544222
GENE-E12-
00152-02
ECCS Suction Strainer Hydraulic Sizing Report
Revision 3
Calculations
SMNH-97-017
Methodology for Sizing Emergency Core Cooling Suction
Strainers
Revision 0
Corrective Action
Documents
CR 11194608
NRC Resident Walkdown comments for Unit 1 drywell
07/12/2025
Corrective Action
Documents
Resulting from
Inspection
CR 11195202
NRC Drywell Closeout Final Walkdown
07/14/2025
Procedures
Normal Plant Shutdown
Version
2.10
CR 11177318
DWFD leakage threshold reached
05/11/2025
CR 11180177
Drywell Floor Drain Inleakage
05/20/2025
CR 11180957
Drywell Floor Drain Leakage
05/23/2025
CR 11186710
Unit 1 SPDS Node 7 failed to reset
06/12/2025
CR 11205344
34SV-E21-001-1 Procedure Change
08/21/2025
Corrective Action
Documents
CR 11211391
2G IRM did not drive in
09/13/2025
Engineering
Evaluations
TE 1179021
Trending and Monitoring: Drywell Floor Drain Inleakage
05/29/2025
RHR Service Water Pump Operability
07/30/2025
Core Spray Pump Operability
08/21/2025
Core Spray Valve Operability
08/19/2025
Diesel Generator 1B Monthly Test
07/18/2025
AMETEK CHARGER FUNTIONAL TEST
11.0
BETTIS 522C-SR Actuator Corrective Maintenance
07/24/2025
Procedures
NMP-MA-012-
Mechanical Pre-Job Brief
06/05/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
F03
SNC 2500150
DWFD leakage threshold reached
05/12/2025
SNC1037020
Disassemble actuator, replace all seals & damaged parts &
reassemble
07/24/2025
SNC1726122
FLEX PM: FLEX Mode 5 core cooling pump functional test &
inspection
06/05/2025
SNC1789429
1B CS pump CPT (Comprehensive Test) per 34SV-E21-
001-1
08/21/2025
SNC1961892
2B CS Loop Valve Operability 34SV-E21-002-2
08/19/2025
SNC2360002
Functional test for EDG 1A Battery Charger
07/23/2025
SNC2400046-20
FT (Unit 2 IRM Instrument)
09/15/2025
SNC2400047
IRM Instrument functional test
09/15/2025
SNC2599398
Repair Leak on DW Cooler 1T47-B007B (DWFD Leakage)
07/11/2025
SNC2615863
Repair Leak on DW Cooler 1T47-B007A (DWFD Leakage)
07/12/2025
SNC2693294
Perform FT after w/o SNC2693277 is completed for MSIV
1B21F028C
07/14/2025
Work Orders
SNC2789082
2G IRM did not drive in - install jumper
09/14/2025
Procedures
H-LT-AF-CPE-
00112
CPE Design - EHC, RPFT, Oil Temp Cont., Cond Vac, ASD
Version 1.3
Self-Assessments
Emergency Preparedness Observation Form
07/07/2025
Corrective Action
Documents
CR 11195381
Justification for not having a loss of safety function
07/15/2025
71151
Miscellaneous
PI information
report
PI information report July 1, 2024, through June 30, 2025
1.0
Corrective Action
Documents
CRs
11167602, 11167623, 11167644
Level Of Evaluation Checklist
07/16/2025
CR 11184106
U1 HPCI Turbine Stop Valve Cycling
06/03/2025
Corrective Action
Documents
CR 11211383
Unit 2 Reactor Scram
09/13/2025
Procedures
Turbine Control Valve Fast Closure Instrument FT
Version 19.4
Work Orders
SNC2574034
U1 HPCI Turbine Stop Valve Cycling
06/05/2025