IR 05000416/2025003
| ML25345A293 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/22/2025 |
| From: | Douglas Dodson NRC/RGN-IV/DORS/PBC |
| To: | Kapellas B Entergy Operations |
| References | |
| EA-22-104, EA-22-115 IR 2025001 | |
| Download: ML25345A293 (0) | |
Text
December 22, 2025
SUBJECT:
GRAND GULF NUCLEAR STATION AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION - INTEGRATED INSPECTION REPORT 05000416/2025003 AND 07200050/2025001
Dear Brad Kapellas:
On September 30, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Grand Gulf Nuclear Station. On December 1, 2025, the NRC inspectors discussed the results of this inspection with Jason Richardson, General Manager Plant Operations, 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 1, 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.
Three findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Grand Gulf Nuclear Station.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC Resident Inspector at Grand Gulf Nuclear Station.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Douglas E. Dodson II, Chief Reactor Projects Branch C Division of Operating Reactor Safety Docket Nos. 05000416 and 07200050 License No. NPF-29
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000416 and 07200050
License Number:
Report Numbers:
05000416/2025003 and 07200050/2025001
Enterprise Identifier:
I-2025-003-0007 and I-2025-001-0012
Licensee:
Entergy Operations, Inc.
Facility:
Grand Gulf Nuclear Station
Location:
Port Gibson, MS
Inspection Dates:
July 1, 2025, to September 30, 2025
Inspectors:
L. Brookhart, Senior Spent Fuel Storage Inspector
K. Clayton, Senior Operations Engineer
J. Freeman, Resident Inspector
C. Harrington, Senior Operations Engineer
J. Kirkland, Senior Operations Engineer
B. Pannabecker, Resident Inspector
A. Smallwood, Senior Resident Inspector
Approved By:
Douglas E. Dodson II, Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Grand Gulf Nuclear Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Maintain Two Fire Barriers in Room 1A405 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.05 The inspectors identified a Green finding and an associated non-cited violation of License Condition 2.C(41), Fire Protection Program, for the licensees failure to implement all provisions of the approved fire protection program described in Updated Final Safety Analysis Report, Section 9A.5.18.1, and in accordance with licensee procedure 06-OP-SP64-R-0047,
Fire-Rated Assembly Visual Inspection. Specifically, the licensee failed to maintain two fire barriers in room 1A405 functional at all times without adequate compensatory measures.
Manual Reactor Scram Due to Degraded Main Condenser Vacuum Caused by a Circulating Water Pump Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000416/2025003-02 Open/Closed None (NPP)71153 The inspectors noted a self-revealed Green finding associated with the licensees failure to identify a single point vulnerability and take appropriate actions to mitigate the concern in accordance with procedures. Specifically, the licensee failed to identify the circulating water system as a single point vulnerability and to mitigate the concern, which led to operators inserting a manual reactor scram on June 17, 2025, upon the trip of one of two circulating water pumps.
Failure to Institute Effective Corrective Actions to Preclude Repetition Resulting in a Loss of Secondary Containment Integrity Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000416/2025003-03 Open/Closed
[H.5] - Work Management 71153 The inspectors noted a self-revealed Green non-cited violation of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee inadvertently left a secondary containment personnel hatch in an open configuration for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while performing corrective roof maintenance, which rendered the secondary containment inoperable. The same secondary containment personnel hatch was previously left open during roof inspections on April 7, 2016, and April 5, 2018, and these repetitive failures were classified by the licensee as a significant condition adverse to quality, but corrective actions to preclude repetition were ineffective.
Additional Tracking Items
Type Issue Number Title Report Section Status NOV 05000416/2023001-01 Failure to Initiate a Condition Report for a Condition Adverse to Quality EA-22-115 71152A Closed NOV 05000416/2023001-06 Failure to Make a Timely Part 21 Report EA-22-104 71152A Closed LER 05000416/2024-002-00 Automatic Actuation of Reactor Protection System 71153 Closed LER 05000416/2024-003-00 Feedwater Inlet Check Valve Incorrectly Determined Operable 71153 Closed LER 05000416/2025-002-00 Automatic Emergency Diesel Generator Actuation Caused by Grid Transient 71153 Closed LER 05000416/2025-003-00 Manual Reactor Scram Due to Degraded Main Condenser Vacuum Caused by Circulating Water Pump Trip 71153 Closed
PLANT STATUS
Grand Gulf Nuclear Station, Unit 1, began the inspection period at 100 percent rated thermal power (RTP). On September 23, 2025, reactor power was reduced to 53 percent RTP to make repairs to the feedwater heater drain system and execute a rod pattern sequence exchange.
The unit returned to 100 percent RTP on September 26, 2025. On September 27, 2025, power was reduced to 79 percent RTP to conduct a rod pattern adjustment. The unit returned to 100 percent RTP on September 28, 2025, where the unit remained at or near for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)division II diesel generator jacket water cooling on July 29, 2025 (2)instrument air system on September 16, 2025
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the division II standby gas treatment system on August 14, 2025.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
(1)division II diesel generator room, fire zone 1D308, on July 31, 2025 (2)fire penetration CE-111D, separating fire zones 1A406 and 1A417, on August 4, 2025 (3)auxiliary building hallway with safety-related cables, fire zone 1A417, on August 13, 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 announced live fire training exercise on August 11, 2025.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
- (1) division II residual heat removal room cooler degraded flow on August 8, 2025
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered on and the biennial written examinations completed on September 2, 2025.
71111.11B - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Program (IP Section 03.04) (1 Sample)
- (1) Biennial Requalification Written Examinations The inspectors evaluated the quality of the licensed operator biennial requalification written examination administered on August 21, 2025.
Annual Requalification Operating Tests The inspectors evaluated the adequacy of the facility licensees annual requalification operating test.
Administration of an Annual Requalification Operating Test The inspectors evaluated the effectiveness of the facility licensee in administering requalification operating tests required by 10 CFR 55.59(a)(2) and that the facility licensee is effectively evaluating their licensed operators for mastery of training objectives.
Requalification Examination Security The inspectors evaluated the ability of the facility licensee to safeguard examination material, such that the examination is not compromised.
Remedial Training and Re-examinations The inspectors evaluated the effectiveness of remedial training conducted by the licensee and reviewed the adequacy of re-examinations for licensed operators who did not pass a required requalification examination.
Operator License Conditions The inspectors evaluated the licensees program for ensuring that licensed operators meet the conditions of their licenses.
Control Room Simulator The inspectors evaluated the adequacy of the facility licensees control room simulator in modeling the actual plant and for meeting the requirements contained in 10 CFR 55.46.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during a power reduction and isolation of a high pressure feedwater heater on September 23, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a licensed operator simulator scenario on August 13, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (4 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)high pressure core spray room cooler degraded flow on August 19, 2025 (2)reactor recirculation pump hydraulic power unit fluid leakage and flow control lockup on August 22, 2025 (3)division II standby service water check valve, P41-005B, removal and testing review completed on September 26, 2025
- (4) Condition Report (CR) CR-GGN-2024-04565, eight motor-operated valves identified that need stem nut replacement review completed on September 29, 2025
Aging Management (IP Section 03.03) (1 Sample)
The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:
(1)auxiliary building and control building concrete cracks, CR-GGN-2025-03537, on September 3, 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) CR-GGN-2021-07298, 1Y79 inverter outage, division I and II standby diesel generator protection on July 28, 2025
- (2) CR-GGN-2025-03383, division II standby service water outage window on August 4, 2025
- (3) CR-GGN-2025-03609, static inverter 1Y82 loss of redundancy on August 21, 2025
- (4) CR-GGN-2025-03898, high pressure feedwater heater steam leakage on September 22, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (3 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) CR-GGN-2025-03026, containment spray time delay relay 1E12-K93B on July 9, 2025 (2)division II diesel generator operability on July 18, 2025 (3)division II standby service cooling tower fan D on August 14, 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)standby service water motor-operated valve P41-F160A post-maintenance test, Work Order (WO) 54033873, on July 24, 2025 (2)standby service water to primary service water crosstie valve P41-F125 rebuild retest, WO 54303525, on August 4, 2025 (3)division II diesel generator lube oil sump heater temperature switch, WO 54299687, on August 8, 2025 (4)standby service water cooling tower fan, P41C003D, WO 54306255, retest on August 12, 2025 (5)diesel-driven fire pump B, WO 54174800, on August 27, 2025 (6)division I standby gas treatment, WO 54133101, on September 25, 2025
Surveillance Testing (IP Section 03.01) (3 Samples)
(1)low pressure core injection/residual heat removal subsystem C check valve test, WO 54265954, on July 12, 2025 (2)low pressure core injection/residual heat removal subsystem pump C quarterly functional test, WO 5426591, on July 12, 2025 (3)standby diesel generator 12 functional test on July 18, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)residual heat removal A shutdown cooling motor-operated valve test, WO 53012421, on July 11,
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) July 1, 2024, through June 30, 2025
MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) July 1, 2024, through June 30, 2025
MS08: Heat Removal Systems (IP Section 02.07) (1 Sample)
- (1) 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)control rod drive mechanism bolt chemical composition on August 19, 2025
- (2) Notices of Violation EA-22-104 and EA-22-115 corrective actions on September 2, 2025. The inspectors reviewed the licensees response to NOV
===05000416/2023001-01 and NOV 0500416/2023001-06 and determined that the reasons, corrective actions taken and planned to address recurrence, and the dates when full compliance will be/was achieved for these violations is adequately addressed and captured on the docket.
(3)standby liquid control environmental qualification on September 24, 2025 71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)===
- (1) The inspectors evaluated the secondary containment loss of safety function and licensee response on September 30, 2025.
Event Report (IP Section 03.02) (4 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000416/2024-002-00, "Automatic Actuation of the Reactor Protection System" (Agencywide Documents Access and Management System (ADAMS) Accession No.
ML24149A159). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to foresee and correct and therefore was not reasonably preventable. No violation of NRC requirements was identified.
- (2) LER 05000416/2024-003-00, "Feedwater Inlet Check Valve Incorrectly Determined Operable" (ML24239A788). The circumstances surrounding this LER and a Green NCV are documented in Inspection Report 05000416/2024010-02 under Inspection Results Section 71153. This LER is closed.
- (3) LER 05000416/2025-002-00, "Automatic Emergency Diesel Generator Actuation Caused by Grid Transient" (ML25133A192). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to be foreseen and corrected and therefore was not reasonably preventable.
No performance deficiency nor violation of NRC requirements was identified. This LER is closed.
- (4) LER 05000416/2025-003-00, "Manual Reactor Scram Due to Lowering Condenser Vacuum" (ML25230A099). The inspection conclusions associated with this LER and Green Finding are documented in this report under Inspection Results Section 71153.
This LER is Closed.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60854 - Preoperational Testing of an Independent Spent Fuel Storage Installation Determine by direct observation and independent evaluation whether the licensee had developed, implemented, and evaluated preoperational testing activities to safely load spent fuel from the site's spent fuel pool to the independent spent fuel storage installation (ISFSI) and safely retrieve spent fuel from the ISFSI and transfer it back to the licensee's spent fuel pool.
Preoperational Testing of an ISFSI===
- (1) Grand Gulf Nuclear Station is a 10 CFR Part 72 general licensee in accordance with 10 CFR 72.210. The licensee originally selected the Holtec International HI-STORM 100 Storage System (Certification of Compliance 72-1014) to store spent fuel at the site's ISFSI. In 2025, the licensee had elected to use the Holtec International HI-STORM FW Storage System (Certificate of Compliance 72-1032) to continue storing fuel at the ISFSI. The ISFSI contained 44 HI-STORM 100 casks holding 68 spent fuel assemblies and the future casks would be HI-STORM FW casks containing 89 spent fuel assemblies. On June 16-19, 2025, inspectors performed onsite inspections to observe and evaluate preoperational testing and training exercises being conducted by the licensee. These operations are described in Condition 9 of the Certificate of Compliance and are performed by a general licensee prior to the use of the system to store spent fuel assemblies.
During the inspection period, inspectors specifically observed the following demonstrations that were successfully completed by the licensee:
- heavy load lifts to remove the HI-TRAC VW transfer cask and simulated canister from the cask washdown pit to above the HI-STORM overpack for download
- heavy load lifts to transfer the simulated canister from the HI-TRAC VW to the HI-STORM overpack
- removal of the empty transfer cask from the stack-up position
===60855 - Operation of an ISFSI The inspectors performed a review of the licensees ISFSI activities to verify compliance with requirements of the two systems in use at the site. The systems include the Holtec Certificate of Compliance (CoC) 72-1032, License Amendment 7 and the HI-STORM FW Final Safety Analysis Report (FSAR), Revision 12 and the Holtec HI-STORM 100 CoC 72-1014, License Amendment 9, Revision 1 and FSAR, Revision 13. The inspectors reviewed selected procedures, corrective action reports, and records to verify ISFSI operations were compliant with the Certificate's technical specifications, requirements in the FSAR, and NRC regulations.
Operation of an ISFSI===
- (1) Inspectors evaluated the licensees dry cask storage loading operations from July 13-14, and July 21-24, 2025, during an onsite inspection. The Grand Gulf ISFSI is sized to store 44 HI-STORM 100 storage casks and 44 HI-STORM FW storage casks. At the time of the inspection, the ISFSI pad contained a total of 44 HI-STORM 100 storage casks, and the licensee was in the process of loading the first HI-STORM FW cask. The HI-STORM FW casks will each contain a multi-purpose canister with 89 fuel assemblies (MPC-89). The triennial review included observation of site loading activities for the first FW storage cask placed into service and a review of evaluations, changes, calculations, and program implementation related to ISFSI activities since the last NRC ISFSI inspection conducted in September 2021.
During the onsite inspection, the inspectors evaluated and observed the following activities:
- walkdown of the ISFSI pad
- heavy load lifts using the cask handling crane to place the canister lid, while under water in spent fuel pool cask loading pit
- heavy load lifts to remove the HI-TRAC VW transfer cask and loaded MPC from the spent fuel pool
- canister lid welding and non-destructive testing activities on the lid-to-shell weld
- canister hydrostatic test
- canister drying
- canister helium backfill
- canister vent port welding, non-destructive testing, and helium leak testing on the vent ports
- heavy load lifts to remove the HI-TRAC VW transfer cask and loaded MPC from the cask washdown pit to above the HI-STORM overpack for download
- heavy load lifts to transfer the loaded MPC from the HI-TRAC VW to the HI-STORM overpack The inspectors reviewed and evaluated the following documentation during the inspection:
- fuel selection evaluations for the canisters loaded since the last NRC ISFSI inspection (canisters 43 - 45)
- radiation surveys for radiological dose at the owner-controlled boundary to verify compliance with the requirements of 10 CFR 72.104 for years 2021 -
2024
- selected ISFSI-related condition reports issued since the last NRC ISFSI inspection
- quality assurance program implementation, including recent audits, surveillances, receipt inspection, and quality control activities related to ISFSI operations
- compliance to technical specifications for operational surveillance activities and FSAR required annual maintenance activities
- selected ISFSI design changes and associated 72.48 screens and evaluations for ISFSI changes that occurred since September 2021
===60856 - Review of 10 CFR 72.212(b) Evaluations Evaluate the licensee's program implementation for inclusion and use of the HI-STORM FW storage system into the site's existing 10 CFR Part 50 and Part 72 programs. The inspection scope included a review of the licensee's programs for heavy loads, emergency planning, fire protection, quality assurance, radiation protection, and site calculations to verify compliance with 10 CFR 72.212 requirements.
Review of 10 CFR 72.212(b) Evaluations===
- (1) From May through July 2025, the NRC conducted an inspection of Grand Gulf Nuclear Station 10 CFR Part 72 program implementation and 10 CFR 72.212 Report prior to the licensee's first loading campaign using the Holtec FW storage system. Licensee calculations and evaluations were reviewed to verify compliance with requirements of the Certificate of Compliance 72-1032, License Amendment 7, and the HI-STORM FW FSAR, Revision 12. The inspectors reviewed selected procedures, corrective action reports, calculations, and evaluations to verify ISFSI operations were compliant with the Certificate's technical specifications, requirements in the FSAR, and NRC regulations.
The following programs, procedures, and calculations were inspected:
- site-specific analysis that confirmed the licensee met the license conditions for flooding, tornados, lightning, blockage of inlet openings, off-normal temperature requirements, and snow and ice
- evaluation to confirm annual dose equivalent would meet 10 CFR 72.104 and 72.106 requirements
- evaluations under 10 CFR 50.59 for the ISFSI's impact on the reactor facility and compliance to site's heavy load program
- calculations for structural and seismic stability for the transfer cask on the fuel building crane and placement locations within the fuel building
- work orders and procedures that the fuel building's crane meets ASME B30.2 for load testing and annual maintenance requirements
- load testing and non-destructive testing records of special lifting devices used during loading operations
- fuel selection procedures to ensure fuel contents meet license conditions
- analyses to determine maximum weights placed on the cask handling crane during loading operations
- fire and explosion hazards analysis for the ISFSI and walkdown of the heavy haul path and ISFSI to ensure all hazards had been identified and evaluated
- quality assurance program, corrective action program, and implementing procedures for incorporation of FW system related activities
- 10 CFR 72.48 program and associated reviews for identified changes to utilize the FW storage system
- comprehensive review of the site's HI-STORM FW 72.212 report, Revision
INSPECTION RESULTS
Failure to Maintain Two Fire Barriers in Room 1A405 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025003-01 Open/Closed
[H.12] - Avoid Complacency 71111.05 The inspectors identified a Green finding and an associated non-cited violation of License Condition 2.C(41), Fire Protection Program, for the licensees failure to implement all provisions of the approved fire protection program described in Updated Final Safety Analysis Report, Section 9A.5.18.1, and in accordance with licensee procedure 06-OP-SP64-R-0047, Fire-Rated Assembly Visual Inspection. Specifically, the licensee failed to maintain two fire barriers in room 1A405 functional at all times without adequate compensatory measures.
Description:
The Grand Gulf Nuclear Station, Unit 1, auxiliary building is a reinforced concrete structure, which surrounds the lower portion of the cylindrical concrete containment structure.
Many of the auxiliary building rooms, 1A405 included, are formed from a combination of auxiliary building concrete walls and the outer wall of the concrete containment. For differential settlement and earthquake considerations, seismic gaps filled with concrete joint sealant are provided at most locations where the auxiliary building concrete meets the containment outer wall. When in a functional status, these sealant-filled seismic gaps meet the required 3-hour fire rating of the concrete barriers.
On June 23, 2025, the inspectors toured the auxiliary building 166-foot elevation including fire zone 1A405, which contains containment ventilation equipment. The inspectors observed abnormal degradation of the sealing material that serves as part of the 3-hour fire barrier separating fire zone 1A405, which contains safety-related electrical cables, and from the auxiliary building hallway fire zone 1A417, which contains safety-related hydrogen analyzer sample racks and electrical cables. The gaps in the sealing material permitted airflow between the two spaces as evident by feel and the visible agitation of fibrous material in the gaps, which were through-cracks smaller than 1/8 inch in width. The sealing material was cracked and separating from the walls.
The inspectors reported the degraded seal to the licensee control room staff. The licensee determined that the fire-rated assembly was nonfunctional, implemented the required hourly fire watch per Technical Requirements Manual (TRM) 6.2.8, and documented this issue in their corrective action program as CR-GGN-2025-02843.
On September 9, 2025, the inspectors observed cracking and separation from the walls in the sealing material that serves as part of the 3-hour fire barrier separating fire zone 1A405 and fire zone 1A124, which is a blowout shaft for several systems including residual heat removal B and reactor core isolation cooling. Due to the height of the cracking the inspectors were unable to take measurements.
The inspectors reported the degraded seal to the licensee control room staff. The licensee determined that the fire-rated assembly was nonfunctional, implemented the required hourly fire watch per TRM 6.2.8, and documented this issue in their corrective action program as CR-GGN-2025-03945.
As part of the licensees fire protection program, UFSAR, Section 9.5.1.2.2.9, requires that fire barriers are capable of containing the effects of possible fires for the minimum amount of time for which the barrier is rated. The licensees procedure 06-OP-SP64-R-0047, Fire-Rated Assembly Visual Inspection, Revision 122, requires that Exposed surfaces of each fire-rated assembly, including structural steel (AND installed fire proofing material) must be free of apparent changes in appearance OR abnormal degradations. The attachments of procedure 06-OP-SP64-R-0047 are performed on a staggered frequency so that each fire-rated assembly is inspected at least once every 10 years. On December 18, 2023, licensee staff inspected fire zones 1A405 and 1A417 and determined the acceptance criteria were met.
Corrective Actions: The licensee established an hourly fire watch in accordance with the TRM and generated work orders for the repair of the degraded 3-hour fire barriers.
Corrective Action References: CR-GGN-2025-02843, CR-GGN-2025-03945
Performance Assessment:
Performance Deficiency: The licensees failure to implement the fire protection program, as required by License Condition 2.C(41), Fire Protection Program, for fire barriers was a performance deficiency. Specifically, the licensee failed to maintain the 3-hour rating of two fire barriers as required by the fire protection program described in UFSAR, Section 9.5.1.2.2.9, and as implemented through licensee procedure 06-OP-SP64-R-0047, Fire-Rated Assembly Visual Inspection, Revision 122, which states, in part, that each fire-rated assembly, including fire proofing material, must be free of apparent changes in appearance or abnormal degradations. This includes cracks, holes, abrasions, and seal separation from walls and components.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in nonfunctional fire barriers separating fire zone 1A405 from fire zones 1A417 and 1A124. The inspectors determined that this issue is similar to the more than minor criteria of example 2.d in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix F, Fire Protection Significance Determination Process, to determine the findings significance. The inspectors evaluated the degraded seal using the included guidance of Appendix F, Attachment 2, Section 2.03.01, and determined that the degradation did not meet the definition of high degradation, and therefore, concluded that the performance deficiency was of very low safety significance (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. Specifically, individual contributors did not perform a thorough review of the work site and planned activity every time work was performed rather than relying on past successes and assumed conditions during recent walkdowns and inspections, which resulted in the barrier degradation going undetected.
Enforcement:
Violation: License Condition 2.C(41), Fire Protection Program, states, in part, that the plant shall implement and maintain in effect all provisions of the fire protection program as described in the UFSAR. The fire protection program, as described in UFSAR, Section 9.5.1.2.2.9 states, in part, that fire-rated penetration seals are provided, as necessary, to maintain the integrity of fire walls. The licensee established licensee procedure 06-OP-SP64-R-0047, Fire-Rated Assembly Visual Inspection, Revision 122, as the implementing procedure to inspect fire-rated assemblies including fire barriers. Step 5.3.1 of the procedure states, in part, that exposed surfaces of each fire-rated assembly must be free of apparent changes in appearance or abnormal degradation including cracks, holes, abrasions, and seal separation from walls.
Contrary to the above, since June 17, 2025, and September 9, 2025, respectively, exposed surfaces of each fire-rated assembly were not free of apparent changes in appearance or abnormal degradation including cracks, holes, abrasions, or seal separation from walls.
Specifically, since June 17, 2025, the fire barrier separating fire zone 1A405 from the auxiliary building hallway, fire zone 1A417, and since September 9, 2025, the fire barrier separating fire zone 1A405 from the 1A124 blowout shaft, were degraded exhibiting cracks and seal separation from the walls rendering the barriers unable to fulfill their safety function as fire barriers.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Manual Reactor Scram Due to Degraded Main Condenser Vacuum Caused by a Circulating Water Pump Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000416/2025003-02 Open/Closed None (NPP)71153 The inspectors noted a self-revealed Green finding associated with the licensees failure to identify a single point vulnerability and take appropriate actions to mitigate the concern in accordance with procedures. Specifically, the licensee failed to identify the circulating water system as a single point vulnerability and to mitigate the concern, which led to operators inserting a manual reactor scram on June 17, 2025, upon the trip of one of two circulating water pumps.
Description:
On June 17, 2025, Grand Gulf Nuclear Station, Unit 1, was operating at 100 percent RTP. At 4:14 a.m. the control room received an alarm for a circulating water pump A trip. The pump trip reduced cooling water flow to the main condenser, resulting in lowering condenser vacuum. Operators entered the Loss of Condenser Vacuum Off-Normal Event procedure and began lowering reactor power with the goal of maintaining condenser vacuum and permitting continued operation. At approximately 4:26 a.m., the control room received the low vacuum annunciator and the operators, in accordance with the Loss of Condenser Vacuum Off-Normal Event procedure, manually inserted a reactor scram.
Following plant shutdown, investigations determined that the circulating water pump trip was caused by a resistance temperature detector (RTD) failure. Each circulating water pump at Grand Gulf Nuclear Station uses two RTDs to inform a digital motor protection relay. This relay was programmed in a one-out-of-two configuration, whereby the relay would actuate and trip the motor breaker if either of the RTDs indicated a temperature greater than the trip setpoint. Troubleshooting determined that one of the RTDs experienced an open circuit failure, causing an erroneous high temperature indication, and resulted in the relay actuation.
Grand Gulf Nuclear Station uses a single point vulnerability (SPV) review process, proceduralized in procedure EN-DC-175, Single Point Vulnerability Review Process, Revision 5, to identify individual components whose failure could require the plant to reduce power (derate) or manually insert a reactor scram. This review process was conducted in 2006, and the circulating water pumps were classified as Derate SPVs, a classification that was supported by historical evidence from 2002. Prior to the units extended power uprate in 2012, the licensee performed another SPV review and concluded that the increase in heat load on the condenser and circulating water systems would be offset by the added cooling water towers and that the Derate classification was still accurate. The licensee failed to recognize this additional heat load would require a manual reactor scram to be inserted due loss of one circulating water pump. The inspectors noted that Single Point Vulnerability Review Process procedure states, in part, that identified vulnerabilities are to be mitigated using the process outlined in the procedure. Failure to identify the loss of a circulating water pump as a plant trip vulnerability resulted in no mitigating strategy being developed to prevent the plant trip.
In 2019 following a change to procedure EN-DC-175, which removed the Derate classification, the station performed a SPV review, and the circulating water pumps were declassified as SPVs, which removed the Derate classification.
During the 2020 refueling outage (RF22), the licensee implemented engineering change (EC)82408 that replaced the circulating water pumps analog overload relays with digital solid-state relays. This EC considered the effects of a failed RTD but concluded that RTD failures could be detected by new functionality provided by the digital solid-state relays.
Corrective Actions: Following plant shutdown, an engineering team performed a failure mode analysis and concluded the circulating water pump thermal overload RTDs were SPVs. The licensee performed WO 54289080 to wire the motor protection relay to a functioning installed spare RTD. The licensee then implemented EC 54289070 and EC 54288806, which changed the thermal overload trip logic for both circulating water pumps to a two-out-of-two configuration, so that a failure of a single RTD would not cause a circulating water pump trip.
Corrective Action References: CR-GGN-2025-02745, WO 54289080, WO 54289063, WO 54289064
Performance Assessment:
Performance Deficiency: The licensees failure to consider the circulating water system as a single point vulnerability, which could cause a unit trip, and develop a mitigating strategy was a performance deficiency. Specifically, procedure EN-DC-175, Single Point Vulnerability Review Process, Revision 12, states in part, that all plant systems should be considered to determine if the plant system is an initiator for a unit trip or turbine trip, and when an SPV is identified, the procedure requires development of a mitigating strategy using the framework of the procedure. Specifically, the failure to implement a mitigating strategy combined with a trip of one of two circulating water pumps on June 17, 2025, caused operators to trip the plant by inserting a manual reactor scram.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment 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, operators inserted a manual reactor scram on June 17, 2025, because the license failed to identify a plant system as an initiator for a unit trip and implement a mitigating strategy.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 1, Initiating Events Screening Questions, Section B, Transient Initiators, the inspectors determined that the performance deficiency caused a reactor trip but did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. There, the issue screened to Green.
Cross-Cutting Aspect: No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Failure to Institute Effective Corrective Actions to Preclude Repetition Resulting in a Loss of Secondary Containment Integrity Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000416/2025003-03 Open/Closed
[H.5] - Work Management 71153 The inspectors noted a self-revealed Green non-cited violation of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee inadvertently left a secondary containment personnel hatch in an open configuration for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while performing corrective roof maintenance, which rendered the secondary containment inoperable. The same secondary containment personnel hatch was previously left open during roof inspections on April 7, 2016, and April 5, 2018, and these repetitive failures were classified by the licensee as a significant condition adverse to quality, but corrective actions to preclude repetition were ineffective.
Description:
Grand Gulf Nuclear Station, Unit 1, secondary containment consists of the reinforced concrete auxiliary building surrounding the lower primary containment and a low-leakage, metal-siding enclosure building surrounding the primary containment above the auxiliary building roofline. The enclosure building has a flat roof, which is periodically inspected and is accessed by internal ladders that lead to an access hatch on the roof.
Access to these ladders requires coordination with other staff when proceeding to the roof.
The roof hatch has a latching mechanism that can only be unlatched from the inside.
On September 4, 2025, plant personnel were working atop the enclosure building performing ongoing roof repairs. Inside the enclosure building on the refueling floor, licensee staff were performing a spent fuel movement evolution. As a planned part of the spent fuel evolution, control room operators manually initiated the standby gas treatment system (SBGT), which is a ventilation system designed to maintain a negative pressure in the atmosphere of the enclosure building during accident conditions. After initiation, operators received the enclosure building negative pressure LO alarm and began investigating. Investigations revealed that the enclosure building roof hatch was unsecured and open by approximately 1 inch. The operators closed and secured the hatch and the alarm cleared. Based on interviews with the plant personnel on the roof and a review of logs for door 1A604, inspectors assessed the hatch was ajar for between 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The licensee documented this condition in the corrective action program as CR-GGN-2025-03819.
In 2016 and 2018 the licensee issued LERs for similar issues where the enclosure building roof hatch was left open for the duration of periodic roof inspections. The 2016 LER, 05000416/2016-003-00, noted that the hatch had likely been left open a minimum of 30 times in the previous 5 years. The licensee performed a root cause evaluation (RCE) to guide corrective actions. The licensee determined that work instructions for the work orders that require personnel to access/egress the secondary containment roof hatch would be revised to ensure the correct control of the hatch and maintenance of secondary containment operability. This event was documented as a Green NCV in NRC Integrated Inspection Report 05000416/2016002 (ML16216A137).
On April 5, 2018, secondary containment was again breached during a routine roof inspection. This event was documented in LER 05000416/2018-002-01. The licensee determined this repeat issue was a significant condition adverse to quality and reopened the RCE from 2016. The licensee noted that the previous corrective action plan did not provide adequate administrative controls and depended heavily on verbal communication to perform the required actions. Inspectors documented an NRC-identified, Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, in NRC Integrated Inspection Report 05000416/2018002 (ML18215A026) for the licensees failure to institute effective corrective actions to preclude repetition of a significant condition adverse to quality.
After the event on September 4, 2025, inspectors reviewed the RCE from 2018 and noted the following corrective actions: 1) update signage, 2) performer talk directly to the Shift Manager, and 3) updates to specific work orders related to routine roof inspections. The inspectors determined that the corrective actions applied in 2016 and 2018 did not address procedures used for corrective maintenance, such as the work being performed on the enclosure building roof in 2025.
Licensee procedure EN-LI-102, Corrective Action Program, Revision 54, defines significant conditions adverse to quality as failures, malfunctions, deficiencies, deviations, defective material, and equipment and nonconformances that adversely affects the safety-related functions of systems, structures, or components (SSCs) deemed significant based on actual or potential consequences to nuclear safety. The procedure states, in part, these conditions (significant conditions adverse to quality) require the cause of the condition be determined and corrective action taken to preclude repetition.
The inspectors reviewed secondary containment design requirements as described in the UFSAR, and observed Section 6.2.3.5 lists all secondary containment doors, main steam tunnel blowout shaft, both residual heat removal blowout shafts, the hatch cover to the radwaste building tunnel, the 185 elevation hatch covers, but is missing the enclosure building roof hatch. Additionally, the inspectors noted that the listed secondary containment openings have indicating alarms, as stated in UFSAR Section 6.2.3.5. The secondary containment enclosure building roof hatch is not listed in Section 6.2.3.5 of the UFSAR and does not have any indications of open or closed position as described in Section 6.2.3.5 of the UFSAR.
Corrective Actions: The licensee closed the enclosure building roof hatch immediately after discovery that it was open, which restored the secondary containment function. The operability of secondary containment was confirmed by operation of the SBGT system and verification of the proper vacuum in the secondary containment prior to commencing with the spent fuel movements. The licensee entered this issue into their corrective action program as CR-GGN-2025-03819. The licensee also stopped all work on the roof of the enclosure building until additional measures could be put in place including a safety standdown to discuss lessons learned with all site personnel.
Corrective Action References: CR-GGN-2025-03819
Performance Assessment:
Performance Deficiency: The licensees failure to implement adequate corrective actions for a significant condition adverse to quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI and licensee procedure EN-LI-102, resulted in inoperability of secondary containment and was a performance deficiency. Specifically, a secondary containment personnel hatch was inadvertently placed in an open configuration for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while performing corrective maintenance on the roof, and the licensee previously identified this condition as a significant condition adverse to quality.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control 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 licensees failure to maintain configuration control of secondary containment while performing roof maintenance resulted in undue inoperability of secondary containment.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions. Using Section D, Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, or SBGT system boiling water reactor.
Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, while performing corrective roof maintenance, the licensee failed to recognize the risk and coordinate between the operations and repair groups to properly maintain secondary containment integrity.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, 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.
Contrary to the above, since April 7, 2016, licensee measures, in the case of a significant condition adverse to quality, failed to assure that the cause of the condition is determined and corrective action taken to preclude repetition. Specifically, licensee corrective actions following 2016 and 2018 failures to maintain control of the secondary containment, a significant condition adverse to quality, did not preclude repetition, and an additional failure to control secondary containment occurred and rendered secondary containment inoperable when a roof hatch was left open during corrective maintenance on the roof.
Enforcement Action: This violation is being treated as an 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 July 29, 2025, the inspectors presented the ISFSI Dry Runs, 72.212 Programs Review, and Triennial Inspections Exit Meeting inspection results to Jeff Hardy, Licensing and Regulatory Assurance Manager, and other members of the licensee staff.
- On August 21, 2025, the inspectors presented the Technical Debrief inspection results to Jayson Kryska, Training Manager, and other members of the licensee staff.
- On September 17, 2025, the inspectors presented the Exit Meeting inspection results to Brad Kapellas, Site Vice President, and other members of the licensee staff.
- On December 1, 2025, the inspectors presented the integrated inspection results to Jason Richardson, General Manager Plant Operations, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
20-S-04-003
20-S-04-004
DFS FW Cask Sealing, Drying, and Backfill
20-S-04-005
DFS FW Cask Transfer, Stack-Up, and Download
60854
Procedures
20-S-04-006
DFS HI-STORM Transport for FW System
Corrective Action
Documents
CR-GGN-2021-
06914, CR-GGN-
21-06971, CR-
GGN-2021-
07387, CR-GGN-
22-08103, CR-
GGN-2023-
226, CR-GGN-
23-13316, CR-
GGN-2023-
14493, CR-GGN-
23-16306, CR-
GGN-2024-
2778, CR-GGN-
25-01736
Various CRs related to dry fuel storage selected for review
during inspection.
Miscellaneous
22 Area
Monitoring Report
Record.pdf, 2023
Area Monitoring
Report Final.pdf,
1st Half 2024
DLR Issue Log -
Final.pdf
TLD records for dose around the ISFSI to meet 10 CFR 2.104 requirements.
60855
Procedures
06-OP-1000-D-
0001-220905,
06-OP-1000-D-
0001-220907,
06-OP-1000-D-
Various technical specification vent surveillances:
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
0001-220909, 06-
OP-1000-D-0001-
20911, 06-OP-
1000-D-0001-
230307, 06-OP-
1000-D-0001-
230309, 06-OP-
1000-D-0001-
230311, 06-OP-
1000-D-0001-
240617, 06-OP-
1000-D-0001-
240619, 06-OP-
1000-D-0001-
240621, 06-OP-
1000-D-0001-
240623.
MPC-744
Fuel selection for Holtec dry cask storage for MPC-744
MPC-745
Fuel selection for Holtec dry cask storage for MPC-745
Self-Assessments QS-2021-GGNS-
030, QA-20-2022-
GGNS-01, QA-
20-2024-GGNS-
QA audits reviewed during inspection.
Holtec DFS Upgrade - HI-STORM FW System Transition
HI-2135869
Site-specific tornado missile analysis for HI-STORM FW
system
HI-2220125
Cask Handling Weights at Grand Gulf Nuclear Station
HI-2220220
Thermal Evaluation of HI-TRAC VW in Cask Washdown Pit
at Grand Gulf
60856
Calculations
HI-2220307-R3,
HI-2220286-R2,
Seismic analysis and static/floor loading analysis for
TC/MPC in pool., in the cask washdown pit, Stackup, rail
Various
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
HI-2220304-R2,
HI-2220313-R2,
HI-2220416-R2,
HI-2220525-R3.
car/low profile trailer, Cask Transporter
HI-2220355
Evaluation of Transporter Fire for HI-STORM 100S Version
B and HI-STORM FW Version F at Grand Gulf
HI-2220590
CoC Dose Evaluation for Grand Gulf Nuclear Station
HI-2220882
Thermal Inertia Calculation for Grand Gulf
Tornado Depressurization and Missile Evaluation of the
1A319A Door
RRTI-3128-0005
LPT Haul Path Evaluation
RRTI-3128-0007
HI-TRAN Seismic Stability on Haul Path with HI-STORM
Engineering
Evaluations
RRTI-3128-0008
Thermal Evaluation of Cask Spacing at Grand Gulf Nuclear
Station
COC 3128121A
Certificate of conformance for canister download slings
Final Safety
Analysis Report
GGNS HI-
STORM 100 10 CFR 72.212
Evaluation Report
Licensing basis document, Docket 72-0050
GGNS HI-
Evaluation Report
Licensing basis document, Docket 72-0050
HI-STORM FW
Certificate of Compliance
Miscellaneous
VI-0045-2023-
GA-2
Purchase Specification for Grand Gulf HI-TRAN Cask
Transporter
20-S-04-017
DFS FW Special Lifting Device Inspections and
Recertification
Corrective Action Program
Procedures
GGNS-CS-20
Standard for heavy loads and special lifting devices
safety-related
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
HI-2220209
Fuel Loading Plan for Grand Gulf Nuclear Station
Work Orders
WO-GGN-
2933594, WO-
WO-GGN-
2972297, WO-
WO-GGN-
53002633.
Work orders to verify annual maintenance/testing of single-
failure proof crane:
Procedures
EN-FAP-EP-010
Severe Weather Response
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
23-15615, 2023-15782, 2024-06007, 2025-00164,
25-01869, 2025-03091, 2022-11160, 2023-00912,
23-01057, 2023-02338, 2023-02342, 2023-14224
04-1-01-P53-1
Instrument Air System
Procedures
04-1-01-T48-1
System Operating Instructions, Standy Gas Treatment
Work Orders
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
25-03533, 2025-02950
Drawings
M1864
Blockouts & Penetrations Auxiliary Building EL. 166'-0" Area-
UNIT-1
10-S-03-7
Fire Protection Training Program
Misc Equip Area - 1A417
Fire Plans
DG-03
Div II Diesel Generator Room1D308
06-OP-SP64-R-
0047
Fire Rated Assembly Visual Inspection
2
Procedures
06-OP-SP64-R-
0049
Fire Rated Sealed Penetrations Visual Inspection
113
Work Orders
WO 54007806, 54104568, 50294444
Corrective Action
CR-GGN-YYYY-
25-00221, 2025-03031
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Documents
NNNN
Work Orders
WO 285171, 54253917, 54221025
Miscellaneous
IP 71111.11 table
03.03-1
Examination Results
09/02/2025
ACA-CR-GGN-
25-00785
ACA on Reactor SCRAM
2/16/2025
CR-GGN-2025-
00619
2/15/2025
CR-GGN-2025-
00823
Low Vacuum SCRAM loss of power
2/17/2025
DR-2018-0005
Simulator Discrepancy on RPV level
01/15/2018
DR-2024-00201
11/10/2024
DR-2025-00008
11/10/2024
Corrective Action
Documents
DR-2025-00049
2/15/2025
25 LORT Written Exam Sample Plan
25
25 GGNS Annual Exam JPM and Scenario Exam
Selection
25
GSES-LOR-
AEX07
Simulator Scenario 7
07/1/2025
GSES-LOR-
AEX08
Simulator Scenario 8
07/1/2025
GSES-LOR-
AEX11
Simulator Scenario 11
06/5/2025
GSES-LOR-
AEX14
Simulator Scenario 14
06/5/2025
JPMs - plant,
simulator, admin
Week 5 (total 6)
JPMs - plant,
simulator, admin
Week 6 (total 6)
SBT-Scenario 3
25
NRC IOL Exam Scenario 3
08/19/2025
Simulator
Differences List
Simulator Differences List
08/19/2025
Miscellaneous
Written Exam
Week 5
06/17/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
EN-OP-115-06
Activation and Deactivation of Licenses, and Maintaining
Active License Status
Operations Training Program Description
Simulator Configuration Control
Licensed Operator Requalification Annual and Biennial
Exam Development
Procedures
TQF-202-SBT
Scenario Based Testing checklist
Procedures
05-S-01-EP-3M1-
Primary Containment Control Modes 1-3
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
24-05323, 2024-05333, 2024-05373, 2024-06818,
25-01574, 2025-02670, 2025-02699, 2025-03304,
23-14897, 2023-15202, 2023-16406, 2023-16468,
23-16644, 2023-16708, 2023-17283, 2023-17391,
24-00053, 2024-00945, 2024-02252, 2024-02324,
24-02550, 2024-02983, 2024-03030, 2024-03411,
24-03994, 2024-06228, 2024-06297, 2025-00489,
25-00802, 2025-01197, 2025-02735, 2025-03429,
25-03523, 2022-08220, 2025-03537
Procedures
ER-GGN-1997-
0761-000
Resolution of GGCR1997-0841-00 cracks in concrete / cmu
walls in the auxiliary building and control building
Work Orders
WO 54077920, 54092426, 00588135
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
25-03383, 2025-03400, 2025-03433, 2025-03524,
21_07298, 2025-03609
Engineering
Changes
54126351
Evaluation of RICT Durations for Grand Gulf Nuclear Station
Work Orders
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
25-03026, 2025-03084
Work Orders
WO 265072, 54242709
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
25-03433, 2025-03045
Procedures
06-OP-1P75-
0002
STANDBY DIESEL GENERATOR (SDG) 12 FUNCTIONAL
TEST
147
Work Orders
WO 53012421, 52685649, 54033873, 54265917, 54303525,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
299687, 54306255, 54174800, 54133101
71151
Procedures
Regulatory Performance Indicator Process
Corrective Action
Documents
CR-YYYY-NNNN
23-17187, 2023-00721, 2023-01768, 2022-09773,
22-10127, 2022-13421, 2025-03721
Engineering
Changes
54084435
NDE Reports
007N9897
Grand Gulf Control Rod Drive Mechanism Cap Screw
Evaluation
Work Orders
WO 243709, 54246673, 54243711, 52985006, 00551962
Corrective Action
Documents
CR-GGN-YYYY-
NNNN
24-02140, 2024-03247