IR 05000416/2025001: Difference between revisions
StriderTol (talk | contribs) StriderTol Bot change |
StriderTol (talk | contribs) StriderTol Bot change |
||
| (4 intermediate revisions by the same user not shown) | |||
| Line 84: | Line 84: | ||
At the beginning of the inspection period, Grand Gulf Nuclear Station, Unit 1, was operating at 88 percent rated thermal power (RTP) at the request of the electric grid operators as a result of transmission line damage from storms in Mississippi that occurred on December 29, 2024. After repairs to power distribution lines were completed, the station achieved 100 percent RTP on January 4, 2025. On February 15, 2025, a reactor SCRAM was manually inserted by the operators, due to a loss of main condenser vacuum. The unit restarted on February 23, 2025, and achieved 100 percent RTP on February 27, 2025. A rod pattern adjustment resulted in a power reduction to 78 percent RTP and a subsequent return to 100 percent RTP on February 28, 2025. Reactor power was reduced to 91 percent RTP on March 23, 2025, to support control rod drive hydraulic control unit maintenance with a subsequent return to 100 percent RTP that same day. On March 29, 2025, a reduction in power to 69 percent RTP was conducted to support a rod pattern sequence exchange. On March 29, 2025, the unit returned to 100 percent RTP where it remained at or near for the rest of the inspection period. | At the beginning of the inspection period, Grand Gulf Nuclear Station, Unit 1, was operating at 88 percent rated thermal power (RTP) at the request of the electric grid operators as a result of transmission line damage from storms in Mississippi that occurred on December 29, 2024. After repairs to power distribution lines were completed, the station achieved 100 percent RTP on January 4, 2025. On February 15, 2025, a reactor SCRAM was manually inserted by the operators, due to a loss of main condenser vacuum. The unit restarted on February 23, 2025, and achieved 100 percent RTP on February 27, 2025. A rod pattern adjustment resulted in a power reduction to 78 percent RTP and a subsequent return to 100 percent RTP on February 28, 2025. Reactor power was reduced to 91 percent RTP on March 23, 2025, to support control rod drive hydraulic control unit maintenance with a subsequent return to 100 percent RTP that same day. On March 29, 2025, a reduction in power to 69 percent RTP was conducted to support a rod pattern sequence exchange. On March 29, 2025, the unit returned to 100 percent RTP where it remained at or near for the rest of the inspection period. | ||
== INSPECTION SCOPES== | ==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. | 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== | ==REACTOR SAFETY== | ||
==71111.01 - Adverse Weather Protection== | ==71111.01 - Adverse Weather Protection== | ||
===Impending Severe Weather Sample (IP Section 03.02) (1 Sample)=== | ===Impending Severe Weather Sample (IP Section 03.02) (1 Sample)=== | ||
| Line 107: | Line 107: | ||
: (2) The inspectors evaluated external flooding mitigation protections related to condition report (CR) GGN-2025-00410, which documents a sinkhole located adjacent to a design-basis flood drainage channel. | : (2) The inspectors evaluated external flooding mitigation protections related to condition report (CR) GGN-2025-00410, which documents a sinkhole located adjacent to a design-basis flood drainage channel. | ||
==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 reactor startup preparations and control rod drive mechanism venting and operation verification on February 20, 2025. | : (1) The inspectors observed and evaluated licensed operator performance in the control room during reactor startup preparations and control rod drive mechanism venting and operation verification on February 20, 2025. | ||
| Line 119: | Line 119: | ||
: (1) RISC-3 categorization of and 10 CFR 50.69 screening for division II standby diesel generator components on February 7, 2025 | : (1) RISC-3 categorization of and 10 CFR 50.69 screening for division II standby diesel generator components on February 7, 2025 | ||
==71111.13 - Maintenance Risk Assessments and Emergent Work Control== | |||
===Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)=== | |||
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed: | 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)standby service water division II and standby diesel generator division II maintenance on February 3, 2025 (2)engineered safety feature 11 transformer maintenance on March 19, 2025 (3)residual heat removal division I on March 26, 2025 | (1)standby service water division II and standby diesel generator division II maintenance on February 3, 2025 (2)engineered safety feature 11 transformer maintenance on March 19, 2025 (3)residual heat removal division I on March 26, 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: | The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments: | ||
(1)standby service water pump division II pump seal leakage, CR-GGN-2025-00222, on January 16, 2025 (2)residual heat removal division II jockey pump issues related to system venting after maintenance, CR-GGN-2025-00249, on January 23, 2025 (3)standby service water valve, P41-F201B, division II standby diesel generator cooler vent packing leakage, CR-GGN-2025-00637, on February 10, 2025 (4)standby diesel generator division II shroud bolt installation, CR-GGN-2025-0995, on February 24, 2025 (5)auxiliary building roof membrane leakage, CR-GGN-2025-00883, on March 5, 2025 | (1)standby service water pump division II pump seal leakage, CR-GGN-2025-00222, on January 16, 2025 (2)residual heat removal division II jockey pump issues related to system venting after maintenance, CR-GGN-2025-00249, on January 23, 2025 (3)standby service water valve, P41-F201B, division II standby diesel generator cooler vent packing leakage, CR-GGN-2025-00637, on February 10, 2025 (4)standby diesel generator division II shroud bolt installation, CR-GGN-2025-0995, on February 24, 2025 (5)auxiliary building roof membrane leakage, CR-GGN-2025-00883, on March 5, 2025 | ||
| Line 138: | Line 138: | ||
: (1) The inspectors evaluated available outage activities following the loss of condenser vacuum and loss of 16BB2 and 16BB4 load centers SCRAM on February 15, 2025. | : (1) The inspectors evaluated available outage activities following the loss of condenser vacuum and loss of 16BB2 and 16BB4 load centers SCRAM on February 15, 2025. | ||
==71111.24 - Testing and Maintenance of Equipment Important to Risk== | |||
Post-Maintenance Testing (PMT) (IP Section 03.01)=== | The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality: | ||
===Post-Maintenance Testing (PMT) (IP Section 03.01) (5 Samples)=== | |||
(1)residual heat removal valve E12-F004B post-maintenance testing, WO 53034421, on January 17, 2025 (2)standby diesel generator division II crankshaft deflection and thrust measurements, WO 53029853, on February 19, 2025 (3)standby diesel generator division II hot adjustment of lifters, WO 53027366, on February 20, 2025 (4)transformer 16BB2 replacement thermography, WO 54133324, on March 18, 2025 | (1)residual heat removal valve E12-F004B post-maintenance testing, WO 53034421, on January 17, 2025 (2)standby diesel generator division II crankshaft deflection and thrust measurements, WO 53029853, on February 19, 2025 (3)standby diesel generator division II hot adjustment of lifters, WO 53027366, on February 20, 2025 (4)transformer 16BB2 replacement thermography, WO 54133324, on March 18, 2025 | ||
: (5) SCRAM time testing for control rods after maintenance, WO 54189810, on March 29, 2025 | : (5) SCRAM time testing for control rods after maintenance, WO 54189810, on March 29, 2025 | ||
| Line 170: | Line 171: | ||
===Inspection Review (IP Section 02.01 - 02.04) (1 Sample)=== | ===Inspection Review (IP Section 02.01 - 02.04) (1 Sample)=== | ||
The inspectors evaluated the licensee's preliminary scenario for the March 18, 2025, biennial emergency preparedness exercise, which was submitted January 16, 2025, to determine whether the proposed scenario would provide adequate opportunities for the licensee's emergency preparedness organization to demonstrate the key emergency preparedness functions. The inspectors consulted with evaluators from the Federal Emergency Management Agency to determine whether the proposed scenario provided adequate opportunities for participating offsite emergency response organizations to achieve their evaluation objectives. | The inspectors evaluated the licensee's preliminary scenario for the March 18, 2025, biennial emergency preparedness exercise, which was submitted January 16, 2025, to determine whether the proposed scenario would provide adequate opportunities for the licensee's emergency preparedness organization to demonstrate the key emergency preparedness functions. The inspectors consulted with evaluators from the Federal Emergency Management Agency to determine whether the proposed scenario provided adequate opportunities for participating offsite emergency response organizations to achieve their evaluation objectives. | ||
: (1) The inspectors reviewed and evaluated the proposed scenario for the 2025 biennial emergency plan exercise and communicated comments to the licensee on February 19, 2025, as required by Inspection Procedure | : (1) The inspectors reviewed and evaluated the proposed scenario for the 2025 biennial emergency plan exercise and communicated comments to the licensee on February 19, 2025, as required by Inspection Procedure 71114.08. | ||
== | ==RADIATION SAFETY== | ||
==71124.01 - Radiological Hazard Assessment and Exposure Controls== | |||
===Radiological Hazard Assessment (IP Section 03.01) (1 Sample)=== | |||
: (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards. | : (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards. | ||
| Line 216: | Line 217: | ||
: (1) three declared pregnant worker evaluations | : (1) three declared pregnant worker evaluations | ||
== OTHER ACTIVITIES - BASELINE== | ==OTHER ACTIVITIES - BASELINE== | ||
71151 - Performance Indicator Verification The inspectors verified licensee performance indicator submittals listed below: | |||
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)=== | |||
===IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (1 Sample)=== | |||
: (1) January 1 through December 31, 2024 | : (1) January 1 through December 31, 2024 | ||
| Line 237: | Line 238: | ||
This is a new NRC performance indicator, introduced in NEI 99-02, Revision 8 (ML24331A114). Licensees began collecting data for this performance indicator on January 1, 2025. Therefore, at the time of inspection, there was no quarterly data compiled and submitted to the NRC as of yet. Verifications will take place on later inspections. | This is a new NRC performance indicator, introduced in NEI 99-02, Revision 8 (ML24331A114). Licensees began collecting data for this performance indicator on January 1, 2025. Therefore, at the time of inspection, there was no quarterly data compiled and submitted to the NRC as of yet. Verifications will take place on later inspections. | ||
===71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)=== | ===71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (1 Sample)=== | ||
{{IP sample|IP=IP 71153|count=1}} | {{IP sample|IP=IP 71153|count=1}} | ||
The inspectors evaluated the following licensee event reports (LERs): | The inspectors evaluated the following licensee event reports (LERs): | ||
: (1) LER 05000416/2024-004-00, High Pressure Core Spray Over Frequency Relay Trip (ML24326A363). The circumstances surrounding this LER and a Green, non-cited violation are documented in the Inspection Results section of this report. This LER is closed. | : (1) LER 05000416/2024-004-00, High Pressure Core Spray Over Frequency Relay Trip (ML24326A363). The circumstances surrounding this LER and a Green, non-cited violation are documented in the Inspection Results section of this report. This LER is closed. | ||
== INSPECTION RESULTS== | ==INSPECTION RESULTS== | ||
Failure to Transfer Licensed Radioactive Material to an Authorized Licensee Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000416/2025001-01 Open/Closed Not Applicable 71124.01 The NRC identified a Severity Level IV non-cited violation of 10 CFR 30.3(a) associated with Grand Gulf Nuclear Stations transfer of their licensed byproduct material to the Entergy Central Calibration Facility. Specifically, the Central Calibration Facility was not authorized or licensed to receive, possess, own, or use the licensed radioactive material. | Failure to Transfer Licensed Radioactive Material to an Authorized Licensee Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000416/2025001-01 Open/Closed Not Applicable 71124.01 The NRC identified a Severity Level IV non-cited violation of 10 CFR 30.3(a) associated with Grand Gulf Nuclear Stations transfer of their licensed byproduct material to the Entergy Central Calibration Facility. Specifically, the Central Calibration Facility was not authorized or licensed to receive, possess, own, or use the licensed radioactive material. | ||
====Description:==== | |||
From approximately calendar year 2000, Entergy Corporation, the parent company of the Grand Gulf Nuclear Station, assumed operation of the radiation instrument Central Calibration Facility adjacent to the Grand Gulf Nuclear Station. The Central Calibration Facility performs calibrations for the Entergy fleet sites: Arkansas Nuclear One, Grand Gulf Nuclear Station, River Bend Station, and Waterford-3. Based on an internal audit, between November 2021 and October 2023, the Central Calibration Facility performed 58 portable instrument calibrations for the four Entergy fleet sites. | From approximately calendar year 2000, Entergy Corporation, the parent company of the Grand Gulf Nuclear Station, assumed operation of the radiation instrument Central Calibration Facility adjacent to the Grand Gulf Nuclear Station. The Central Calibration Facility performs calibrations for the Entergy fleet sites: Arkansas Nuclear One, Grand Gulf Nuclear Station, River Bend Station, and Waterford-3. Based on an internal audit, between November 2021 and October 2023, the Central Calibration Facility performed 58 portable instrument calibrations for the four Entergy fleet sites. | ||
| Line 258: | Line 259: | ||
Specifically, Grand Gulf Nuclear Station had inappropriately transferred byproduct material, which required a specific license to possess, to an unlicensed entity. | Specifically, Grand Gulf Nuclear Station had inappropriately transferred byproduct material, which required a specific license to possess, to an unlicensed entity. | ||
Corrective Actions: The licensee reviewed their Quality Assurance Program Manual to determine what changes in organizational structure have occurred, if any, and what changes are required to bring the licensed radioactive material back into the control of the NRC licensee, Grand Gulf Nuclear Station. | ====Corrective Actions:==== | ||
The licensee reviewed their Quality Assurance Program Manual to determine what changes in organizational structure have occurred, if any, and what changes are required to bring the licensed radioactive material back into the control of the NRC licensee, Grand Gulf Nuclear Station. | |||
Corrective Action References: CR-GGN-2024-06745, CR-HQN-2024-01270, and CR-GGN-2025-01218 | Corrective Action References: CR-GGN-2024-06745, CR-HQN-2024-01270, and CR-GGN-2025-01218 | ||
====Performance Assessment:==== | |||
The NRC determined this violation was associated with a minor ROP performance deficiency and is best assessed using traditional enforcement. | The NRC determined this violation was associated with a minor ROP performance deficiency and is best assessed using traditional enforcement. | ||
====Enforcement:==== | |||
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. | The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. | ||
Severity: This issue was determined to be a Severity Level IV, non-cited violation, consistent with NRCs Enforcement Policy, dated August 23, 2024, as informed by materials operations example 6.3.d.9, in which a licensee fails to seek NRC approval, when required, before changing the location where licensed activities are being conducted or where licensed material is being stored that has little or no radiological or programmatic significance, and all other safety and security requirements have been met. In this case, the inspectors determined that the Central Calibration Facility had appropriate security measures in place for the transferred licensed byproduct material stored in the facility. | Severity: This issue was determined to be a Severity Level IV, non-cited violation, consistent with NRCs Enforcement Policy, dated August 23, 2024, as informed by materials operations example 6.3.d.9, in which a licensee fails to seek NRC approval, when required, before changing the location where licensed activities are being conducted or where licensed material is being stored that has little or no radiological or programmatic significance, and all other safety and security requirements have been met. In this case, the inspectors determined that the Central Calibration Facility had appropriate security measures in place for the transferred licensed byproduct material stored in the facility. | ||
Violation: Title 10 CFR 30.3(a) states, in part, no person shall manufacture, produce, transfer, receive, acquire, own, possess, or use byproduct material except as authorized in a specific or general license issued in accordance with the regulations in this chapter. | ====Violation:==== | ||
Title 10 CFR 30.3(a) states, in part, no person shall manufacture, produce, transfer, receive, acquire, own, possess, or use byproduct material except as authorized in a specific or general license issued in accordance with the regulations in this chapter. | |||
Contrary to the above, from calendar year 2000 through January 2025, the Grand Gulf Nuclear Station transferred byproduct material without authorization in a specific or general license. Specifically, NRC inspectors determined that the licensee, Grand Gulf Nuclear Station, had transferred byproduct material to the Entergy Fleet Central Calibration Facility in calendar year 2000. This calibration facility was not authorized to receive, possess, own or use byproduct material because they did not possess a specific or general license. | Contrary to the above, from calendar year 2000 through January 2025, the Grand Gulf Nuclear Station transferred byproduct material without authorization in a specific or general license. Specifically, NRC inspectors determined that the licensee, Grand Gulf Nuclear Station, had transferred byproduct material to the Entergy Fleet Central Calibration Facility in calendar year 2000. This calibration facility was not authorized to receive, possess, own or use byproduct material because they did not possess a specific or general license. | ||
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. | ====Enforcement Action:==== | ||
This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. | |||
Failure to Correct a Condition Adverse to Quality for the High-Pressure Core Spray Over Frequency Relay Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025001-02 Open/Closed | Failure to Correct a Condition Adverse to Quality for the High-Pressure Core Spray Over Frequency Relay Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025001-02 Open/Closed | ||
| Line 280: | Line 284: | ||
Conservative Bias 71153 The inspectors reviewed a self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, from July 2023 to September 2024, the licensee failed to correct three conditions adverse to quality associated with high-pressure core spray system over frequency relay reliability, as evidenced by surveillance issues in July 2023 and September 2024, which caused the high-pressure core spray system to be inoperable. | Conservative Bias 71153 The inspectors reviewed a self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, from July 2023 to September 2024, the licensee failed to correct three conditions adverse to quality associated with high-pressure core spray system over frequency relay reliability, as evidenced by surveillance issues in July 2023 and September 2024, which caused the high-pressure core spray system to be inoperable. | ||
====Description:==== | |||
On June 28, 2023, during a division III planned maintenance outage, the 17AC-181, high-pressure core spray (HPCS) pump over frequency trip relay (the relay), was found to be out of calibration and would not calibrate, as documented in CR-GGN-2023-14155. The relay was replaced with a new relay taken from on-site stores. | On June 28, 2023, during a division III planned maintenance outage, the 17AC-181, high-pressure core spray (HPCS) pump over frequency trip relay (the relay), was found to be out of calibration and would not calibrate, as documented in CR-GGN-2023-14155. The relay was replaced with a new relay taken from on-site stores. | ||
| Line 309: | Line 313: | ||
The inspectors noted that Technical Specification 3.5.1 requires each emergency core cooling system injection/spray subsystem, which includes HPCS, to be operable in MODES 1, 2, and 3. If HPCS is inoperable in MODES 1, 2, or 3, Technical Specification 3.5.1, Condition B requires reactor core isolation cooling (RCIC) to be verified operable within 1 hour and HPCS restored to operable within 14 days or the allowable risk informed completion time. If these actions are not completed, the reactor must be in MODE 3 within 12 hours. Considering the licensee inappropriately determined the HCPS pump was operable in July 2023 when it would have tripped instantly on LOP/LOCA, preventing it from meeting its intended safety function to inject water, this represents a condition prohibited by technical specifications. | The inspectors noted that Technical Specification 3.5.1 requires each emergency core cooling system injection/spray subsystem, which includes HPCS, to be operable in MODES 1, 2, and 3. If HPCS is inoperable in MODES 1, 2, or 3, Technical Specification 3.5.1, Condition B requires reactor core isolation cooling (RCIC) to be verified operable within 1 hour and HPCS restored to operable within 14 days or the allowable risk informed completion time. If these actions are not completed, the reactor must be in MODE 3 within 12 hours. Considering the licensee inappropriately determined the HCPS pump was operable in July 2023 when it would have tripped instantly on LOP/LOCA, preventing it from meeting its intended safety function to inject water, this represents a condition prohibited by technical specifications. | ||
Corrective Actions: The licensee corrected the equipment issue on September 29, 2024, by completing WO 53019672 and replacing the failed relay and installing a matching capacitor. | ====Corrective Actions:==== | ||
The licensee corrected the equipment issue on September 29, 2024, by completing WO 53019672 and replacing the failed relay and installing a matching capacitor. | |||
Post-maintenance testing conducted as part of WO 53019672 confirmed system operability was restored. | Post-maintenance testing conducted as part of WO 53019672 confirmed system operability was restored. | ||
| Line 315: | Line 320: | ||
Corrective Action References: This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2024-05318. | Corrective Action References: This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2024-05318. | ||
====Performance Assessment:==== | |||
Performance Deficiency: The licensee failed to adequately evaluate and correct conditions adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the HPCS motor over frequency relay, which led to the HPCS pump failing surveillances in July 2023 and September 2024. Specifically, in July 2023 the licensee failed to identify and correct conditions adverse to quality associated with the over frequency relay: | Performance Deficiency: The licensee failed to adequately evaluate and correct conditions adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the HPCS motor over frequency relay, which led to the HPCS pump failing surveillances in July 2023 and September 2024. Specifically, in July 2023 the licensee failed to identify and correct conditions adverse to quality associated with the over frequency relay: | ||
: (1) the degraded and nonconforming relay had been installed for at least 35 years, which was well past its expected service life of 8 years per EPRI guidance, and condition monitoring revealed degradation in July 2023, but the licensee reinstalled the original relay and failed to take appropriate action; | : (1) the degraded and nonconforming relay had been installed for at least 35 years, which was well past its expected service life of 8 years per EPRI guidance, and condition monitoring revealed degradation in July 2023, but the licensee reinstalled the original relay and failed to take appropriate action; | ||
| Line 321: | Line 326: | ||
: (3) when the relay failed on time delay, the licensee evaluated the relay as operable and failed to recognize the relay degradation could cause the HPCS over frequency relay to trip instantaneously. As a result, the licensee placed the relay back in service in July 2023 and it tripped instantaneously during surveillance testing in September 2024. | : (3) when the relay failed on time delay, the licensee evaluated the relay as operable and failed to recognize the relay degradation could cause the HPCS over frequency relay to trip instantaneously. As a result, the licensee placed the relay back in service in July 2023 and it tripped instantaneously during surveillance testing in September 2024. | ||
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human 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 (i.e., core damage). Specifically, failing to ensure the operability of the HPCS over frequency relay caused the system to be inoperable from July 2, 2023, through September 24, 2024. | ====Screening:==== | ||
The inspectors determined the performance deficiency was more than minor because it was associated with the Human 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 (i.e., core damage). Specifically, failing to ensure the operability of the HPCS over frequency relay caused the system to be inoperable from July 2, 2023, through September 24, 2024. | |||
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that a detailed risk evaluation was required for this finding because the finding involved the loss of a single-train technical specification system longer than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation using SPAR model version 8.82 with SAPHIRE version 8.2.11. The analyst determined that the performance deficiency resulted in a degraded condition that caused the HPCS system to be unavailable to perform its required safety function if powered from the division III diesel generator in the event of a LOP. Based on the performance and maintenance history associated with the 17AC-181 relay described above, the analyst determined that an exposure time of 1 year would be applicable. While the degraded condition of the relay precluded successful operation of the HPCS system in the event of a LOP, the analyst evaluated whether a subsequent manual start of the HPCS system powered from the division III diesel generator would have been possible. The analyst reviewed licensee operations procedures and determined that, given an initial trip of the system generated by the degraded 17AC-181 relay, subsequent manual operation of the system would be achievable during a design-basis event involving a LOP. Accordingly, the analyst determined that a recovery action to restore the HPCS function following a LOP condition using procedural operator actions would be applicable with a failure probability of 3E-2 in the conditional risk case. The analyst modified the fault tree HCS-ACT (HPCS Actuation and Control) to include a loss of power to the 17ac bus as being a failure mode for HPCS in the conditional case, with this recovery action available to preclude that failure mode. The analyst also set the basic event FLX-XHE-XE-ELAP (Operators Fail to Declare ELAP When Beneficial) to a failure probability of 1E-2 to reflect credit for the use of diverse and flexible (FLEX) coping strategies. With these assumptions, the analyst quantified the SPAR model for all internal and external events except fire events to determine a total increase in average annual core damage frequency (delta-CDF) of 3.96E-7/yr. Dominant sequences included: | ====Significance:==== | ||
The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that a detailed risk evaluation was required for this finding because the finding involved the loss of a single-train technical specification system longer than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation using SPAR model version 8.82 with SAPHIRE version 8.2.11. The analyst determined that the performance deficiency resulted in a degraded condition that caused the HPCS system to be unavailable to perform its required safety function if powered from the division III diesel generator in the event of a LOP. Based on the performance and maintenance history associated with the 17AC-181 relay described above, the analyst determined that an exposure time of 1 year would be applicable. While the degraded condition of the relay precluded successful operation of the HPCS system in the event of a LOP, the analyst evaluated whether a subsequent manual start of the HPCS system powered from the division III diesel generator would have been possible. The analyst reviewed licensee operations procedures and determined that, given an initial trip of the system generated by the degraded 17AC-181 relay, subsequent manual operation of the system would be achievable during a design-basis event involving a LOP. Accordingly, the analyst determined that a recovery action to restore the HPCS function following a LOP condition using procedural operator actions would be applicable with a failure probability of 3E-2 in the conditional risk case. The analyst modified the fault tree HCS-ACT (HPCS Actuation and Control) to include a loss of power to the 17ac bus as being a failure mode for HPCS in the conditional case, with this recovery action available to preclude that failure mode. The analyst also set the basic event FLX-XHE-XE-ELAP (Operators Fail to Declare ELAP When Beneficial) to a failure probability of 1E-2 to reflect credit for the use of diverse and flexible (FLEX) coping strategies. With these assumptions, the analyst quantified the SPAR model for all internal and external events except fire events to determine a total increase in average annual core damage frequency (delta-CDF) of 3.96E-7/yr. Dominant sequences included: | |||
station blackout events with failure of HPCS and either RCIC or FLEX, as well as LOP events with failures of HPCS, reactor core isolation cooling, control rod drives, and failure of manual depressurization. The analyst determined that best available information to estimate the applicable increase in risk due to fire events would be based on the analysts review of results from the licensees fire probabilistic risk assessment model. Based on this review, the analyst determined that a delta-CDF of 5.08E-9/yr due to fire events would be applicable. The analyst determined that risk attributable to large early release frequency was not a dominant metric in the significance determination for this finding. Based on a total delta-CDF of 4.01E-7/yr, the analyst concluded that the finding was of very low safety significance (Green). | station blackout events with failure of HPCS and either RCIC or FLEX, as well as LOP events with failures of HPCS, reactor core isolation cooling, control rod drives, and failure of manual depressurization. The analyst determined that best available information to estimate the applicable increase in risk due to fire events would be based on the analysts review of results from the licensees fire probabilistic risk assessment model. Based on this review, the analyst determined that a delta-CDF of 5.08E-9/yr due to fire events would be applicable. The analyst determined that risk attributable to large early release frequency was not a dominant metric in the significance determination for this finding. Based on a total delta-CDF of 4.01E-7/yr, the analyst concluded that the finding was of very low safety significance (Green). | ||
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision-making practices that emphasize prudent choices over those that are simply allowable. The proposed action is determined to be safe to proceed, rather than unsafe to stop. Specifically, managers did not take timely action to address degraded HPCS conditions commensurate with their safety significance and ensure that conservative assumptions were used when determining how to proceed when degraded HPCS conditions were identified by the licensee. | ====Cross-Cutting Aspect:==== | ||
H.14 - Conservative Bias: Individuals use decision-making practices that emphasize prudent choices over those that are simply allowable. The proposed action is determined to be safe to proceed, rather than unsafe to stop. Specifically, managers did not take timely action to address degraded HPCS conditions commensurate with their safety significance and ensure that conservative assumptions were used when determining how to proceed when degraded HPCS conditions were identified by the licensee. | |||
====Enforcement:==== | |||
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance's are promptly identified and corrected. | Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance's are promptly identified and corrected. | ||
| Line 335: | Line 343: | ||
Contrary to the above, between July 2023 and September 2024, measures were not established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances were promptly identified and corrected. Specifically, between July 2023 and September 2024, the licensee failed to correct three conditions adverse to quality that caused HPCS and its over frequency relay to be inoperable and the over frequency relay tripped during surveillances in July 2023 and September 2024. This also represents a condition prohibited by Technical Specification 3.5.1. | Contrary to the above, between July 2023 and September 2024, measures were not established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances were promptly identified and corrected. Specifically, between July 2023 and September 2024, the licensee failed to correct three conditions adverse to quality that caused HPCS and its over frequency relay to be inoperable and the over frequency relay tripped during surveillances in July 2023 and September 2024. This also represents a condition prohibited by Technical Specification 3.5.1. | ||
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. | ====Enforcement Action:==== | ||
This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. | |||
== EXIT MEETINGS AND DEBRIEFS== | ==EXIT MEETINGS AND DEBRIEFS== | ||
The inspectors verified no proprietary information was retained or documented in this report: | The inspectors verified no proprietary information was retained or documented in this report: | ||
* On March 11, 2025, the inspectors presented the occupational radiation safety inspection results to Brad Kapellas, Site Vice President, and other members of the licensee staff. | * On March 11, 2025, the inspectors presented the occupational radiation safety inspection results to Brad Kapellas, Site Vice President, and other members of the licensee staff. | ||
Latest revision as of 00:18, 23 February 2026
| ML26041A340 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/13/2026 |
| From: | Douglas Dodson NRC/RGN-IV/DORS/PBC |
| To: | Kapellas B Entergy Operations |
| Shared Package | |
| ML26041A334 | List: |
| References | |
| IR 2025001 | |
| Download: ML26041A340 (0) | |
Text
February 13, 2026
SUBJECT:
GRAND GULF NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000416/2025001 - REVISED
Dear Brad Kapellas:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Grand Gulf Nuclear Station. On April 15, 2025, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of the inspection were originally documented in NRC Inspection Report 05000416/2025001, issued on May 13, 2025 (Agencywide Documents Access and Management Systems [ADAMS] Accession No.
During our review of calendar year 2025 inspection reports, we determined that some emergency preparedness inspections that were completed were inadvertently left out of the original report. These inspections did not identify any additional findings or violations. This report is being reissued to document completion of those inspections. Changes were made to Sections 71114.01, 71114.04, and 71151. No changes were made to the results documented in the original report. The revised inspection report is enclosed.
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 Signed by Dodson, Douglas on 02/13/26 Docket No. 05000416 License No. NPF-29
Enclosure:
As stated
Inspection Report
Docket Number:
05000416
License Number:
Report Number:
Enterprise Identifier:
I-2025-001-0005
Licensee:
Entergy Operations, Inc.
Facility:
Grand Gulf Nuclear Station
Location:
Port Gibson, MS
Inspection Dates:
January 1, 2025, to March 31, 2025
Inspectors:
L. Carson, Senior Health Physicist
R. Kopriva, Senior Project Engineer
J. O'Donnell, Senior Health Physicist
B. Pannabecker, Resident Inspector
A. Smallwood, Senior Resident Inspector
P. Elkmann, Senior Emergency Preparedness 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 Transfer Licensed Radioactive Material to an Authorized Licensee Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000416/2025001-01 Open/Closed Not Applicable 71124.01 The NRC identified a Severity Level IV non-cited violation of 10 CFR 30.3(a) associated with Grand Gulf Nuclear Stations transfer of their licensed byproduct material to the Entergy Central Calibration Facility. Specifically, the Central Calibration Facility was not authorized or licensed to receive, possess, own, or use the licensed radioactive material.
Failure to Correct a Condition Adverse to Quality for the High-Pressure Core Spray Over Frequency Relay Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025001-02 Open/Closed
[H.14] -
Conservative Bias 71153 The inspectors reviewed a self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, from July 2023 to September 2024, the licensee failed to correct three conditions adverse to quality associated with high-pressure core spray system over frequency relay reliability, as evidenced by surveillance issues in July 2023 and September 2024, which caused the high-pressure core spray system to be inoperable.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000416/2024-004-00 High Pressure Core Spray Over Frequency Relay Trip 71153 Closed
PLANT STATUS
At the beginning of the inspection period, Grand Gulf Nuclear Station, Unit 1, was operating at 88 percent rated thermal power (RTP) at the request of the electric grid operators as a result of transmission line damage from storms in Mississippi that occurred on December 29, 2024. After repairs to power distribution lines were completed, the station achieved 100 percent RTP on January 4, 2025. On February 15, 2025, a reactor SCRAM was manually inserted by the operators, due to a loss of main condenser vacuum. The unit restarted on February 23, 2025, and achieved 100 percent RTP on February 27, 2025. A rod pattern adjustment resulted in a power reduction to 78 percent RTP and a subsequent return to 100 percent RTP on February 28, 2025. Reactor power was reduced to 91 percent RTP on March 23, 2025, to support control rod drive hydraulic control unit maintenance with a subsequent return to 100 percent RTP that same day. On March 29, 2025, a reduction in power to 69 percent RTP was conducted to support a rod pattern sequence exchange. On March 29, 2025, the unit returned to 100 percent RTP where it remained at or near for the rest 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
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk significant systems from impending severe weather during a high wind advisory from the National Weather Service on March 3, 2025.
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)standby service water division II following maintenance on February 6, 2025 (2)diesel generator division II following maintenance on February 25, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 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)standby service water pump building division II, fire zone 2M110, following maintenance on February 6, 2025 (2)lower cable spreading room, fire zone OC402, 148' elevation of the control building on February 10, 2025 (3)auxiliary building to primary containment seals on elevations 166' and 185', fire zones 1A420 and 1A519, on February 16, 2025 (4)diesel generator room division II, fire zone 1D303, following maintenance on February 25, 2025 (5)engineered safety feature switchgear room, fire zone 1A308, following maintenance on February 27, 2025
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated flooding mitigation protections in the manhole, MH-21, for safety-related cables and sump pump operation located near standby service water basins, as documented in work order (WO) 54156336.
- (2) The inspectors evaluated external flooding mitigation protections related to condition report (CR) GGN-2025-00410, which documents a sinkhole located adjacent to a design-basis flood drainage channel.
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 reactor startup preparations and control rod drive mechanism venting and operation verification on February 20, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator simulator scenarios on March 27, 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) RISC-3 categorization of and 10 CFR 50.69 screening for division II standby diesel generator components on February 7, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
(1)standby service water division II and standby diesel generator division II maintenance on February 3, 2025 (2)engineered safety feature 11 transformer maintenance on March 19, 2025 (3)residual heat removal division I on March 26, 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)standby service water pump division II pump seal leakage, CR-GGN-2025-00222, on January 16, 2025 (2)residual heat removal division II jockey pump issues related to system venting after maintenance, CR-GGN-2025-00249, on January 23, 2025 (3)standby service water valve, P41-F201B, division II standby diesel generator cooler vent packing leakage, CR-GGN-2025-00637, on February 10, 2025 (4)standby diesel generator division II shroud bolt installation, CR-GGN-2025-0995, on February 24, 2025 (5)auxiliary building roof membrane leakage, CR-GGN-2025-00883, on March 5, 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)standby service water pump shaft run out, engineering change (EC) 79246, on February 5, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated available outage activities following the loss of condenser vacuum and loss of 16BB2 and 16BB4 load centers SCRAM on February 15, 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) (5 Samples)
(1)residual heat removal valve E12-F004B post-maintenance testing, WO 53034421, on January 17, 2025 (2)standby diesel generator division II crankshaft deflection and thrust measurements, WO 53029853, on February 19, 2025 (3)standby diesel generator division II hot adjustment of lifters, WO 53027366, on February 20, 2025 (4)transformer 16BB2 replacement thermography, WO 54133324, on March 18, 2025
- (5) SCRAM time testing for control rods after maintenance, WO 54189810, on March 29, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
(1)reactor core isolation cooling/residual heat removal steam line flow functional test channel B, WO 54200399, on January 8, 2025 (2)drywell purge system division I isolation valve E61-F005A monthly test, WO 53028242, on January 21, 2025
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)residual heat removal valve E12-F003B stroke time testing, WO 53034425, on January 17, 2025
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) FLEX building No.1, portable generator test, WO 53028121, on January 14, 2025
71114.01 - Exercise Evaluation
Inspection Review (IP Section 02.01-02.11) (1 Partial)
(1)
(Partial)
The inspectors evaluated the biennial emergency plan exercise conducted on March 18, 2025. However, as of the end of the 1st Quarter 2025, the licensee had not completed its critique process for the exercise. As such, the inspectors had not completed the full evaluation of the exercise and their critique process. The inspectors will complete the inspection in the 2nd Quarter, and the results of the inspection will be documented in NRC Inspection Report 05000416/2025002.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors reviewed the list of changes made to the licensee's emergency plan and emergency plan implementing procedures between June 1, 2024, and March 1, 2025. The inspectors did not identify any changes that required NRC review according to Inspection Procedure 71114.04.
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
(1)blue team emergency preparedness drill on March 4, 2025
71114.08 - Exercise Evaluation - Scenario Review
Inspection Review (IP Section 02.01 - 02.04) (1 Sample)
The inspectors evaluated the licensee's preliminary scenario for the March 18, 2025, biennial emergency preparedness exercise, which was submitted January 16, 2025, to determine whether the proposed scenario would provide adequate opportunities for the licensee's emergency preparedness organization to demonstrate the key emergency preparedness functions. The inspectors consulted with evaluators from the Federal Emergency Management Agency to determine whether the proposed scenario provided adequate opportunities for participating offsite emergency response organizations to achieve their evaluation objectives.
- (1) The inspectors reviewed and evaluated the proposed scenario for the 2025 biennial emergency plan exercise and communicated comments to the licensee on February 19, 2025, as required by Inspection Procedure 71114.08.
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)surveys of potentially contaminated material leaving the radiologically controlled area (RCA)
(2)workers exiting the RCA during routine plant operations
Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
(1)condensate full-flow filter - B filter changes out and support work, radiological work permit (RWP) 2025-1099 Task-1, Revision 01, Radiation Area (RA)
(2)condensate full-flow filter - B filter changes out and support work, RWP 2025-1099 Task-6, Revision 01, High Radiation Area (HRA) and Locked High Radiation Area (LHRA)
(3)low-pressure feed water heater helium leak investigation, RWP 2025-1054 Task-6
- (4) Shepherd calibration source vendor repair, RWP 2024-1052 (5)general tours and inspection, RWP 2025-1005
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following HRAs and very high radiation areas:
- (1) HRA - pre-coat filter room on the 133-foot elevation area-5 of the turbine building
- (2) HRA - condenser area-6 of the turbine building 93-foot elevation
- (3) HRA - offgas recombiner and reheater room on the 93-foot elevation area-1 of the auxiliary building
- (4) HRA - pipe chase/tunnel room on the 87-foot elevation area-1 of the auxiliary building
- (5) HRA - condenser retubing area on the 87-foot elevation area-1 of the turbine building Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (2 Samples)
The inspectors evaluated the following internal dose assessments:
(1)containment worker intake (inhalation) on March 6, and 10, 2024 (2)condenser bay worker intake (ingestion) on March 24, 2024
Special Dosimetric Situations (IP Section 03.04) (1 Sample)
The inspectors evaluated the following special dosimetric situations:
- (1) three declared pregnant worker evaluations
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicator submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (1 Sample)
- (1) January 1 through December 31, 2024
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (1 Sample)
- (1) January 1 through December 31, 2024
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
- (1) January 1 through December 31, 2024
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) January 1 through December 31, 2024 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) January 1 through December 31, 2024
EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)
- (1) April 1 through December 31, 2024 EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)
- (1) April 1 through December 31, 2024 EP04: Emergency Response Facility and Equipment Readiness (ERFER) (IP Section 02.14)
This is a new NRC performance indicator, introduced in NEI 99-02, Revision 8 (ML24331A114). Licensees began collecting data for this performance indicator on January 1, 2025. Therefore, at the time of inspection, there was no quarterly data compiled and submitted to the NRC as of yet. Verifications will take place on later inspections.
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000416/2024-004-00, High Pressure Core Spray Over Frequency Relay Trip (ML24326A363). The circumstances surrounding this LER and a Green, non-cited violation are documented in the Inspection Results section of this report. This LER is closed.
INSPECTION RESULTS
Failure to Transfer Licensed Radioactive Material to an Authorized Licensee Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000416/2025001-01 Open/Closed Not Applicable 71124.01 The NRC identified a Severity Level IV non-cited violation of 10 CFR 30.3(a) associated with Grand Gulf Nuclear Stations transfer of their licensed byproduct material to the Entergy Central Calibration Facility. Specifically, the Central Calibration Facility was not authorized or licensed to receive, possess, own, or use the licensed radioactive material.
Description:
From approximately calendar year 2000, Entergy Corporation, the parent company of the Grand Gulf Nuclear Station, assumed operation of the radiation instrument Central Calibration Facility adjacent to the Grand Gulf Nuclear Station. The Central Calibration Facility performs calibrations for the Entergy fleet sites: Arkansas Nuclear One, Grand Gulf Nuclear Station, River Bend Station, and Waterford-3. Based on an internal audit, between November 2021 and October 2023, the Central Calibration Facility performed 58 portable instrument calibrations for the four Entergy fleet sites.
In January 2025, inspectors reviewed the Central Calibration Facility's Radioactive Source Inventory List and noted that at least 15 radioactive sources, listed as "licensed," required a specific license to possess and were housed at the Central Calibration Facility for the purpose of instrument calibrations. Inspectors determined that all 15 of these inventoried radioactive sources were transferred from the Grand Gulf Nuclear Station to the Central Calibration Facility and were not being inventoried under the Grand Gulf Nuclear Station, Unit 1, license (NPF-29).
Inspectors determined that the Central Calibration Facility is operated in accordance with Entergy fleet Procedures EN-RP-317, Central Calibration Facility, Revision 2, and EN-FAP-RP-007, Operation of the [Radiation Protection] RP Central Calibration Facility, Revision 3. Both fleet procedures state that the Entergy Fleet Radiation Protection Manager is responsible for the overall implementation of this program. Both procedures also state that each Entergy sites RP Manager is responsible for returning instruments to the Central Calibration Facility for calibration. Based on this, inspectors determined that the Central Calibration Facility staff responsible for access and control of the transferred radioactive sources stationed at the Central Calibration Facility do not fall within the organizational structure of the Grand Gulf Nuclear Station license.
Inspectors questioned under what license the Central Calibration Facility was operating. Per the licensees response to NRC questions, the licensee assumed the Central Calibration Facility was part of their operating license since the facility is located within the owner control area of the site.
The inspectors concluded that the transfer of licensed radioactive material to the Central Calibration Facility from the GGNS was not done in accordance with NRC regulations.
Specifically, Grand Gulf Nuclear Station had inappropriately transferred byproduct material, which required a specific license to possess, to an unlicensed entity.
Corrective Actions:
The licensee reviewed their Quality Assurance Program Manual to determine what changes in organizational structure have occurred, if any, and what changes are required to bring the licensed radioactive material back into the control of the NRC licensee, Grand Gulf Nuclear Station.
Corrective Action References: CR-GGN-2024-06745, CR-HQN-2024-01270, and CR-GGN-2025-01218
Performance Assessment:
The NRC determined this violation was associated with a minor ROP performance deficiency and is best assessed using traditional enforcement.
Enforcement:
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.
Severity: This issue was determined to be a Severity Level IV, non-cited violation, consistent with NRCs Enforcement Policy, dated August 23, 2024, as informed by materials operations example 6.3.d.9, in which a licensee fails to seek NRC approval, when required, before changing the location where licensed activities are being conducted or where licensed material is being stored that has little or no radiological or programmatic significance, and all other safety and security requirements have been met. In this case, the inspectors determined that the Central Calibration Facility had appropriate security measures in place for the transferred licensed byproduct material stored in the facility.
Violation:
Title 10 CFR 30.3(a) states, in part, no person shall manufacture, produce, transfer, receive, acquire, own, possess, or use byproduct material except as authorized in a specific or general license issued in accordance with the regulations in this chapter.
Contrary to the above, from calendar year 2000 through January 2025, the Grand Gulf Nuclear Station transferred byproduct material without authorization in a specific or general license. Specifically, NRC inspectors determined that the licensee, Grand Gulf Nuclear Station, had transferred byproduct material to the Entergy Fleet Central Calibration Facility in calendar year 2000. This calibration facility was not authorized to receive, possess, own or use byproduct material because they did not possess a specific or general license.
Enforcement Action:
This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Correct a Condition Adverse to Quality for the High-Pressure Core Spray Over Frequency Relay Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2025001-02 Open/Closed
[H.14] -
Conservative Bias 71153 The inspectors reviewed a self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, from July 2023 to September 2024, the licensee failed to correct three conditions adverse to quality associated with high-pressure core spray system over frequency relay reliability, as evidenced by surveillance issues in July 2023 and September 2024, which caused the high-pressure core spray system to be inoperable.
Description:
On June 28, 2023, during a division III planned maintenance outage, the 17AC-181, high-pressure core spray (HPCS) pump over frequency trip relay (the relay), was found to be out of calibration and would not calibrate, as documented in CR-GGN-2023-14155. The relay was replaced with a new relay taken from on-site stores.
The new relay was calibrated with the previously installed capacitor from the failed relay.
On July 1, 2023, during surveillance testing of the system with the new relay, the HPCS pump tripped on over frequency. The licensee performed troubleshooting of the 17AC-181 relay, removed it from the bus, and calibrated the relay. The licensee determined the trip time was drifting and the relay was unable to be calibrated. This issue was repeatable, and the licensee determined it indicated a nonconforming part or a gap in the procedure instructions.
The licensee initially proposed the cause of the failure as infant mortality and sent the failed relay and its associated capacitor for analysis to determine if a degraded condition was present.
A new relay and capacitor set was received from another station and was calibrated in accordance with Grand Gulf Nuclear Station procedure. The calibration was successful; however, the licensee determined that the originally installed capacitor could not be removed from the system without risk of loss of power to the division III electrical bus due to current plant configuration. This led to the licensees decision to install the new calibrated relay with the already installed original capacitor. The licensee documented this in CR-GGN-2023-14202, which states, to prevent risk associated with losing the 17 bus and to exit the [limiting condition for operation (LCO)] in a timely manner, the original capacitor was used with the new relay. The HPCS pump tripped again on over frequency during subsequent testing.
The inspectors noted the vendor manual states that when external capacitors are furnished with their over frequency relays, they are identified by means of serial numbers to ensure that each relay that is to be installed is provided with the same auxiliaries with which it was calibrated in the factory. This means that, if the relays are calibrated on-site, they should be installed with the same capacitor with which they were calibrated.
On July 2, 2023, the licensee made the decision to re-install the original relay (the relay that would not calibrate) with the original capacitor (the capacitor that remained in the division III bus). The intent of the decision was to prevent the pump trip even though the relay would be out of tolerance on time delay (failed the trip time requirement of 1.25 +/- 0.09 seconds with a trip time of 20.12 seconds). The licensee performed an operability determination in CR-GGN-2023-14202 to justify this decision and assumed that the pump would start and run at a higher frequency.
On September 24, 2024, during division III loss of offsite power/loss of coolant accident (LOP/LOCA) testing, the HPCS pump tripped, as documented in CR-GGN-2024-05318. The trip occurred at 62.4 Hz, below the setpoint of 63 Hz +/- 0.5 Hz. CR-GGN-2024-05335 documents the physical condition of the 17AC-181 over frequency trip relay noting: 1) one of the coils had evidence of insulation distortion that indicated a potential for coil overheating with smells indicating that overheating occurred, 2) the timing cam spring had evidence of discoloration, 3) the timing cam backstop bar was not seating properly, 4) the cam initiation trip contact showed evidence the cam was out of time, and 5) when bench testing the relay, tapping on the bench caused the relay to trip. A new relay and matching capacitor were then installed for the 17AC-181 relay and the LOP/LOCA surveillance was again performed on September 27, 2024, with the HPCS pump running successfully powered from the division III electrical bus being supplied with power from the standby diesel generator division III.
Post-event, the capacitor was tested in accordance with recommendations from the Electric Power Research Institute (EPRI), and it failed three out of four test parameters. It failed capacitance, dielectric absorption, and leakage current. Failure analysis by the vendor determined the shelf life state of the relay was not in accordance with vendor specifications, specifically the relay may revert to a tripped state in a deenergized condition.
The licensee performed a root cause evaluation and determined the direct cause of failure was due to age related degradation of the relay, capacitor, or both. The relay had been in service for over 35 years. EPRI guidance recommends replacement of this type of relay at 8 years to prevent age related failures. Additionally, the licensee determined the root cause was due to having an unnecessary over frequency relay in the HPCS system, and the licensee identified multiple other contributing causes: not testing the capacitor supplied with the original relay, incorrect calibration and installation of the relay (not a matching pair), and not replacing a relay that was at least 35 years old.
The inspectors noted that during evaluation of this deficient condition in July 2023, the licensee did not recognize that degradation of the time delay mechanism springs and contacts could lead to an instantaneous trip of the relay causing a loss of safety function.
Specifically, the inspectors noted that the quarterly surveillance procedure for the HPCS pump, Procedure 06-OP-1E22-Q-0005, HPCS Quarterly Functional Test, Revision 136, is conducted using offsite power aligned to the division III bus. During the test, offsite power is at a stable 60 Hz and would not cause the over frequency relay to begin its timing or initiate its trip function. However, during LOP/LOCA surveillance testing using Procedure 06-OP-1P81-R-0001, HPCS Diesel Generator Function Test, Test No. 2 - LOP and LOCA Safety Related, Revision 139, as was the case in September 2024 when the 17AC-181 relay tripped instantaneously, the division III bus is deenergized and the division III standby diesel generator rapidly starts to supply power to the division III electrical bus and HPCS pump. A rapid start of the standby diesel generator during simulated LOP/LOCA conditions will cause an electrical bus transient that tests the functionality of the over frequency relay.
Considering all of the above, the inspectors assessed that the licensees operability evaluation and decision-making to re-install the original degraded relay that could not be calibrated was not justifiable and demonstrated inadequate decision-making. Additionally, the degraded and nonconforming relay had been installed for at least 35 years, and condition monitoring revealed degradation in July 2023, but the licensee reinstalled the original relay and failed to take appropriate action. Finally, the licensee failed to calibrate, install, and test the relay and external capacitor as a matched pair in July 2023.
The inspectors noted that Technical Specification 3.5.1 requires each emergency core cooling system injection/spray subsystem, which includes HPCS, to be operable in MODES 1, 2, and 3. If HPCS is inoperable in MODES 1, 2, or 3, Technical Specification 3.5.1, Condition B requires reactor core isolation cooling (RCIC) to be verified operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and HPCS restored to operable within 14 days or the allowable risk informed completion time. If these actions are not completed, the reactor must be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Considering the licensee inappropriately determined the HCPS pump was operable in July 2023 when it would have tripped instantly on LOP/LOCA, preventing it from meeting its intended safety function to inject water, this represents a condition prohibited by technical specifications.
Corrective Actions:
The licensee corrected the equipment issue on September 29, 2024, by completing WO 53019672 and replacing the failed relay and installing a matching capacitor.
Post-maintenance testing conducted as part of WO 53019672 confirmed system operability was restored.
Corrective Action References: This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2024-05318.
Performance Assessment:
Performance Deficiency: The licensee failed to adequately evaluate and correct conditions adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the HPCS motor over frequency relay, which led to the HPCS pump failing surveillances in July 2023 and September 2024. Specifically, in July 2023 the licensee failed to identify and correct conditions adverse to quality associated with the over frequency relay:
- (1) the degraded and nonconforming relay had been installed for at least 35 years, which was well past its expected service life of 8 years per EPRI guidance, and condition monitoring revealed degradation in July 2023, but the licensee reinstalled the original relay and failed to take appropriate action;
- (2) the relay needed to be calibrated and installed with a matched external capacitor, but the licensee failed to calibrate, install, and test the relay and external capacitor as a matched pair in July 2023; and
- (3) when the relay failed on time delay, the licensee evaluated the relay as operable and failed to recognize the relay degradation could cause the HPCS over frequency relay to trip instantaneously. As a result, the licensee placed the relay back in service in July 2023 and it tripped instantaneously during surveillance testing in September 2024.
Screening:
The inspectors determined the performance deficiency was more than minor because it was associated with the Human 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 (i.e., core damage). Specifically, failing to ensure the operability of the HPCS over frequency relay caused the system to be inoperable from July 2, 2023, through September 24, 2024.
Significance:
The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that a detailed risk evaluation was required for this finding because the finding involved the loss of a single-train technical specification system longer than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation using SPAR model version 8.82 with SAPHIRE version 8.2.11. The analyst determined that the performance deficiency resulted in a degraded condition that caused the HPCS system to be unavailable to perform its required safety function if powered from the division III diesel generator in the event of a LOP. Based on the performance and maintenance history associated with the 17AC-181 relay described above, the analyst determined that an exposure time of 1 year would be applicable. While the degraded condition of the relay precluded successful operation of the HPCS system in the event of a LOP, the analyst evaluated whether a subsequent manual start of the HPCS system powered from the division III diesel generator would have been possible. The analyst reviewed licensee operations procedures and determined that, given an initial trip of the system generated by the degraded 17AC-181 relay, subsequent manual operation of the system would be achievable during a design-basis event involving a LOP. Accordingly, the analyst determined that a recovery action to restore the HPCS function following a LOP condition using procedural operator actions would be applicable with a failure probability of 3E-2 in the conditional risk case. The analyst modified the fault tree HCS-ACT (HPCS Actuation and Control) to include a loss of power to the 17ac bus as being a failure mode for HPCS in the conditional case, with this recovery action available to preclude that failure mode. The analyst also set the basic event FLX-XHE-XE-ELAP (Operators Fail to Declare ELAP When Beneficial) to a failure probability of 1E-2 to reflect credit for the use of diverse and flexible (FLEX) coping strategies. With these assumptions, the analyst quantified the SPAR model for all internal and external events except fire events to determine a total increase in average annual core damage frequency (delta-CDF) of 3.96E-7/yr. Dominant sequences included:
station blackout events with failure of HPCS and either RCIC or FLEX, as well as LOP events with failures of HPCS, reactor core isolation cooling, control rod drives, and failure of manual depressurization. The analyst determined that best available information to estimate the applicable increase in risk due to fire events would be based on the analysts review of results from the licensees fire probabilistic risk assessment model. Based on this review, the analyst determined that a delta-CDF of 5.08E-9/yr due to fire events would be applicable. The analyst determined that risk attributable to large early release frequency was not a dominant metric in the significance determination for this finding. Based on a total delta-CDF of 4.01E-7/yr, the analyst concluded that the finding was of very low safety significance (Green).
Cross-Cutting Aspect:
H.14 - Conservative Bias: Individuals use decision-making practices that emphasize prudent choices over those that are simply allowable. The proposed action is determined to be safe to proceed, rather than unsafe to stop. Specifically, managers did not take timely action to address degraded HPCS conditions commensurate with their safety significance and ensure that conservative assumptions were used when determining how to proceed when degraded HPCS conditions were identified by the licensee.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance's are promptly identified and corrected.
Technical Specification 3.5.1, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System, requires, in part, that each ECCS injection/spray subsystem is operable in MODES 1, 2, and 3. If HPCS is inoperable, Condition B requires RCIC to be verified operable by administrative means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and HPCS restored to operable within 14 days or the allowable risk informed completion time. If required action and associated completion time of Condition B is not met, the reactor must be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Contrary to the above, between July 2023 and September 2024, measures were not established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances were promptly identified and corrected. Specifically, between July 2023 and September 2024, the licensee failed to correct three conditions adverse to quality that caused HPCS and its over frequency relay to be inoperable and the over frequency relay tripped during surveillances in July 2023 and September 2024. This also represents a condition prohibited by Technical Specification 3.5.1.
Enforcement Action:
This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report:
- On March 11, 2025, the inspectors presented the occupational radiation safety inspection results to Brad Kapellas, Site Vice President, and other members of the licensee staff.
- On March 24, 2025, the inspectors presented the emergency preparedness exercise scenario review inspection results to Jason Richardson, General Manager Plant Operations, and other members of the licensee staff.
- On April 15, 2025, the inspectors presented the integrated inspection results to Brad Kapellas, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00624, 2025-00785, 2025-00883,
25-00985, 2025-00995
04-1-01-R21-16
System Operation Instruction ESF
Bus 16AB
07-S-22-P75-E001B-1
Standby Diesel Generator Slip Ring/Brush
Inspection and Generator/Roto Megger
Test
Operations Equipment Labeling
Procedures
GGNA-GES-01
General Engineering Standard for A Total
Plant Marking System
Work Orders
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00659, 2025-00884, 2025-00930
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00410
Work Orders
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00935, 2025-00936
Calculations
ENTG-GGNS-00176-CALC-007
GGNS 10 CFR 50.59 Categorization for
the DG System
Procedures
Risk Informed Completion Time
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00553, 2025-01320, 2025-01504,
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-00222, 2025-00249, 2025-00637,
25-00883, 2025-0995
Engineering
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Changes
Engineering
Changes
Procedures
Conduct of Operations
Work Orders
WO 53018121, 53022842, 53027366,
53029853, 53034421, 53034425,
54133324, 54189810, 54200399,
221596, 54243516
Corrective
Action
Documents
CR-GGN-2025-01364, GGN-2025-
01382
Condition Reports generated during the
Inspection
February 22, 2023, Dress Rehearsal
Exercise Evaluation Report
GIN 2023-000059
EOF Mini Drill Critique
GIN 2023-000115
26July2023 Blue Team Exercise Critique
GIN 2023-00096
First Half Semi-Annual Health Physics
Drill Report
GIN 2023-00140
13Sep2023 Red Team Exercise Critique
GIN 2023-00155
25Oct2023 Green Team Exercise Critique
GIN 2024-00010
23 GGNS Onsite Accountability Drill
GIN 2024-00014
Second Half 2023 Semi-Annual Health
Physics Drill
GIN 2024-00128
28Aug2024 Yellow-Red Team Turnover
Drill
GIN 2025-00005
3Dec2024 Onsite Medical Drill with CC
Medical Center and ASAP Ambulance
GIN 2025-00009
20Dec2024 2024 Radiological Monitoring
Drill
GIN 2025-00010
18Dec2024 Second Half 2024 Semi-
Annual Health Physics Drill
GIN 2025-00010
6Jun2024 First Half Semi-Annual Health
Physics Drills
Miscellaneous
GIN 2025-0004
2Dec2024 2024 Onsite Accountability
Drill
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
05-S-02-VI-3
Off Normal Event Procedure, Earthquake
10-S-01-1
Activation of the Emergency Plan, R133,
10/27/2020
10-S-01-12
Radiological Assessment and Protective
Action Recommendations, R50,
11/28/2022
10-S-01-14
Emergency Radiation Monitoring, R26,
5/19/2019
Drills and Exercises, R11, 2/10/2022
Offsite Dose Assessment using the
Unified RASCAL Interface, R4, 4/22/2020
Emergency Notifications, R1, 12/15/2021
Operations, R7, 05/01/2023
Technical Support Center Operations, R8,
05/01/2023
Operations Support Center Operations,
R8, 05/01/2023
EPP 01-02
Grand Gulf Nuclear Station, Emergency
Action Levels Wall Chart, Revision A
Procedures
GIN 2022-00195
Grand Gulf Nuclear Station, Emergency
Action Level Technical Bases, R3
Miscellaneous
List of Changes to the Licensee
Emergency Plan and Emergency Plan
Implementing Procedures, June 1, 2024,
to March 1, 2025
Corrective
Action
Documents
CR-GGN-YYYY-NNNNN
25-01151, 2025-01152, 2025-01153,
25-01154, 2025-01155, 2025-01156,
25-01158, 2025-01159, 2025-01160
GGNS Scenario HH2_R1-12-21
Grand Gulf Nuclear Station, EAL
Technical Bases
Miscellaneous
21-3-16_GGNS_IPX_Exercise
Grand Gulf Nuclear Station #2021-3-16
Emergency Preparedness Ingestion
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Pathway Exercise Scenario
GGNS-NRC-01
Draft Scenario for the March 18, 2025,
Biennial Exercise
GGNS_2023_03_22_NRC_EXERCISE
23 Emergency Preparedness Site
Integrated Drill Scenario
05-S-01-EP-2M1-3
RPV Control Modes 1-3
05-S-01-SAP-1M1-4
RPV Control Modes 1-4
Procedures
10-S-01-12
Radiological Assessment and Protective
Action Recommendations
RWP 2024-1301
ALARA Plan
01/21/2023
RWP 2024-1402
RWP Termination and Post-Job ALARA
Review
04/10/2024
ALARA Plans
RWP 2024-1403
RWP Termination and Post-Job ALARA
Review
05/23/2024
Corrective
Action
Documents
CR-GGN
24-01870, 2024-01885, 2024-02048,
24-02085, 2024-02641, 2024-02756,
24-03181, 2024-03757, 2024-03764,
24-04959, 2024-06107, 2024-06412,
25-00514
Corrective
Action
Documents
Resulting from
Inspection
CR-GGN
25-00556, 2025-01218
Refueling Outage 23 Outage ALARA
Report February 26, 2022, to May 2, 2022
08/09/2022
208' Aux Pool Inventory
Status Board for 208' Auxiliary Building
Pools
2/15/2024
EN-RP-143, Attachment 5
Central Facility Source Inventory
01/13/2025
EN-RP-143, Attachment 5
Central Facility Source Inventory
01/09/2024
EN-RP-143, Attachments 4 & 5
GGNS Leak Test & Source Inventory
11/14/2023
EN-RP-143, Attachments 4 & 5
GGNS Leak Test & Source Inventory
11/11/2024
Miscellaneous
GIN-2014-00024
2013 National Source Tracking System
Reconciliation for NRC License NFP-29
01/29/2014
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Entergy Quality Assurance Program
Manual
RF24
Refueling Outage 24 Outage ALARA
Report
04/18/2024
UCP-2501
Containment Pools Cycle 25
2/06/2025
08-S-01-28
Use and Control of Temporary Shielding
Radiation Worker Expectations
Access Control for Radiologically
Controlled Areas
Radiological Survey Documentation
EN-RP-106-1
Radiological Survey Guidelines
Radiation Protection Posting
ALARA Program
ALARA Program
EN-RP-110-04
Radiation Protection Risk Assessment
Process
EN-RP-110-06
Outage Dose Estimating and Tracking
Radioactive Material Control
EN-RP-121-01
Receipt of Radioactive Material
Radiological Controls for Highly
Radioactive Objects
Air Sampling
Procedures
Conduct of Radiation Protection
Hot Tool Room Area. Elevation 133 U2
11/19/2024
Turbine Building Elevation 166 North
Section with CFFF-B Removal HRA and
LHRA dose rates
2/04/2025
Radiation
Surveys
Turbine Building Elevation 166 North
Section
2/05/2025
24-1052
Shepherd Calibration Source Vendor
Repair
Radiation Work
Permits (RWPs)
24-1301
Planned or Forced Outage-PM's, RT's,
Surveillances, Tours, Inspection, and
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
General Maintenance in the Drywell
24-1403
Reactor Vessel Disassembly and
Reassembly
25-1002
Ops Rounds, Surveillances, Tagouts, and
Tours
25-1004
General Maintenance Activities and
Support Work
25-1005
Tours and Inspections
25-1006
Radwaste Shipping and Processing
25-1099
Condensate Full-Flow Filter (CFFF) - B
Filter change out and support work
Self-
Assessments
Pre-NRC Inspection Self-Assessment
11/13/2024
Main Steam Line Monitor Calibration and
Shephard Calibrator Failure
05/21/2024
Main Steam Line Monitor - D Calibration
01/22/2025
Work Orders
Condensate Full-Flow Filter (CFFF) - B
Filter change out and support work
2/06/2025
24 WBC Long
Calibration of the Mirion FASTSCAN
Whole Body Counting System
07/09/2024
Calibration
Records
24 WBC New
Calibration of the Mirion FASTSCAN
Whole Body Counting System
07/12/2024
Corrective
Action
Documents
CR-GGN-
23-16603, 2023-17118, 2024-01223,
24-01743, 2024-02094, 2024-02386,
24-02530, 2024-04184, 2024-04321,
24-05883
Certificate No. L24-761
Certificate of Accreditation for Teledyne
Brown Engineering for Chemical and
Environmental Testing
10/05/2024
EN-RPT-19-003-R0
Standardization of a Bias Between DLRs
and SRDs in the Entergy Fleet:
Update Based on 2018 Results
10/25/2019
Miscellaneous
Generic Radiation Worker Training
Lesson Plan, Section 3.23
07/23/2020
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
GGN-RPT-19-010-R0
Prospective Evaluation for Grand Gulf
Nuclear Station
08/25/2020
GGN-RPT-21-002R0
Neutron Monitoring at Grand Gulf Nuclear
Station
08/19/2021
GGN-RPT-23-002 R0
Grand Gulf Nuclear Station Alpha / Plant
Characterization Data
11/08/2023
GIN-2024-00064
24-2026 10CFR61 Waste Stream
Analysis for Dry Active Waste
05/08/2024
NVLAP Lab Code: 100518-0
Scope of Accreditation to ISO/IEC
17025:2017
2/31/2024
NVLAP Lab Code: 100518-0
Certificate of Accreditation to ISO/IEC
17025:2017
2/31/2023
Air Sampling
Dosimetry Administration
Personnel Monitoring
Dose Assessment
Prenatal Monitoring
Procedures
Whole Body Counting / In-Vitro Bioassay
Radiation
Surveys
Annual Dosimetry Office
2/08/2024
Pre-NRC Inspection Occupational Dose
Assessment IP 71124.04
11/17/2024
Self-
Assessments
QA-14/15-2023-GGNS-01
Quality Assurance Audit Report:
Combined Radiation Protection (RP) and
Radwaste (RW)
2/11/2023
CR-GGN-YYYY-NNNNN
24-05221, 2024-06703
Corrective
Action
Documents
Condition Report generated during the
Inspection
2nd Qtr EOF 04/10/2024 - Scenario
Documents
2nd Qtr EOF 05/01/2024 - Scenario
Documents
71151
Miscellaneous
3rd Qtr EOF 11/30/2023 (Date for
previous training, but ERO member
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
credited once qualified) - Scenario
Documents
3rd Qtr EOF 05/22/2023 (Date for
previous training, but ERO member
credited once qualified) - Scenario
Documents
1st Qtr 2025 SM/SIM (01/15/2025) -
Scenario Documents
1st Qtr 2025 SM/SIM (02/05/2025) -
Scenario Documents
4th Qtr. PI Data Sheet Occupational
Radiation Safety
Occupational Exposure Control
Effectiveness
01/08/25
4th Qtr. PI Data Sheet Public Radiation
Safety
RETS/ODCM Radiological Effluent
Occurrence
01/25/25
Grand Gulf Emergency Response
Organization Pool and Minimum Staff
Positions
10-S-01-12
Radiological Assessment and Protective
Action Recommendations, R50
11/28/2022
Drills and Exercises, R11
2/10/2022
Offsite Dose Assessment using the
Unified RASCAL Interface, R4
04/22/2020
Emergency Notifications, R1
2/15/2021
Regulatory Performance Indicator
Process, R21
05/30/2023
Regulatory Performance Indicator
Process
EP 01-02
Grand Gulf Nuclear Station, Emergency
Action Levels Wallchart, Revision A
Procedures
GIN 2022-00196
Grand Gulf Nuclear Station, Emergency
Action Level Technical Bases, R3