IR 05000354/2025001

Integrated Inspection Report 05000354/2025001

ML25119A117
Person / Time
Site:	Hope Creek
Issue date:	04/29/2025
From:	Nicole Warnek Division of Operating Reactors
To:	Mcfeaters C Public Service Enterprise Group
References
IR 2025001
Download: ML25119A117 (1)

Text

April 29, 2025

SUBJECT:

HOPE CREEK GENERATING STATION - INTEGRATED INSPECTION REPORT 05000354/2025001

Dear Charles McFeaters:

On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Hope Creek Generating Station. On April 9, 2025, the NRC inspectors discussed the results of this inspection with Eric Larson, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Two findings of very low safety significance (Green) are documented in this report. Both of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Hope Creek Generating Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Nicole S. Warnek, Chief Projects Branch 3 Division of Operating Reactor Safety

Docket No. 05000354 License No. NPF-57

Enclosure:

As stated

Inspection Report

Docket Number:

05000354

License Number:

NPF-57

Report Number:

05000354/2025001

Licensee:

PSEG Nuclear, LLC - N09

Facility:

Hope Creek Generating Station

Location:

Hancocks Bridge, NJ

Inspection Dates:

January 01, 2025 to March 31, 2025

Inspectors:

P. Finney, Senior Resident Inspector

J. Bresson, Resident Inspector

E. Chen, Reactor Inspector

M. Henrion, Senior Project Engineer

S. Mercurio, Emergency Preparedness Inspector

E. Miller, Senior Reactor Inspector

R. Rolph, Senior Health Physicist

Approved By:

Nicole S. Warnek, Chief

Projects Branch 3

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Hope Creek Generating 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

Inadequate Procedure for Calibration of the Voltage and Current Transducers Associated with a 4 Kilovolt (kV) Vital Bus Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000354/2025001-01 Open/Closed None (NPP)71152A A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified when PSEG failed to maintain a calibration procedure for a safety-related 4kV vital bus.

Inadequate Airlock Installation Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000354/2025001-02 Open/Closed None (NPP)71152A A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when PSEG did not provide appropriate quantitative or qualitative acceptance criteria in a containment personnel airlock installation procedure to determine it had been satisfactorily accomplished.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000354/2025001-03 Unresolved Item Regarding the NRC's Grant of a Notice of Enforcement Discretion (NOED) for an EDG under TS 3.8.1.1 71153 Open LER 05000354/2024-001-00 LER 2024-001-00 for Hope Creek Generating Station,

Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT)71153 Closed LER 05000354/2024-001-01 LER 2024-001-01 for Hope Creek Generating Station,

Invalid Primary Containment Integrated Leak Rate As-71153 Closed

Found Test (ILRT)

Supplement

PLANT STATUS

Unit 1 began the inspection period at rated thermal power. On January 11, 2025, the unit was reduced to 70 percent power in support of turbine valve testing, a control rod pattern exchange, and to swap infeeds on non-1E buses. The unit was returned to rated thermal power on January 12, 2025. On January 31, 2025, with two emergency diesel generators (EDGs)inoperable, PSEG commenced a technical specification-required shutdown. Power was reduced to 81 percent before the 'A' EDG was restored to operable status. PSEG raised power and reached 100 percent on February 1, 2025. Unit 1 remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed 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.

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated readiness for impending adverse weather for a cold weather alert between January 21 and 23, 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)

'A' through 'C' EDGs during 'D' EDG unavailability, February 5, 2025 (2)

'A' control room emergency filtration during 'B' control room emergency filtration inlet damper planned maintenance, February 19, 2025

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 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) High-pressure coolant injection pump and turbine room, fire plan 3413, January 13, 2025

(2) High-pressure coolant injection and reactor core isolation coolant battery rooms, fire plan 3512, January 23, 2025

(3) Heating, ventilation, and air conditioning equipment, inverter and battery rooms following temporary heating installed for lowered battery room temperature, fire plan 3562, January 30, 2025

(4) Electrical access area during 'D' EDG lube oil troubleshooting, fire plan 3533, February 5, 2025

(5) Control area heating, ventilation, and air conditioning equipment rooms, fire plan 3563, March 6, 2025

(6) Control equipment mezzanine area during degradation requiring continuous fire watch, fire plan 3542, March 12, 2025

(7) Turbine building, fire plan 3121, March 31, 2025

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the main control room during the failure and inoperability of the 'C' refueling floor exhaust radiation monitor on February 11, 2025.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed and evaluated license operator performance during a simulator evaluation on January 21, 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)

'D' EDG water in lube oil, January 31, 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) Emergent work on the 3B feedwater heater following an unplanned trip on a high-level signal, January 14, 2025

(2) Elevated risk during scheduled maintenance on 'A' EDG, week of January 27, 2025

(3) Elevated risk during emergent 'D' EDG inoperability for water in lube oil, week of February 3, 2025

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1)

'B' filtration, recirculation, and ventilation system train indicating flow while out of service following flow transmitter replacement, January 13, 2025 (2)

'A' station service water crosstie due to a 30 drop per minute leak, January 14, 2025

(3) Turbine auxiliary cooling system inboard supply isolation valve from safety auxiliary cooling system due to its stroke time, January 15, 2025

(4) Reactor core isolation cooling batteries due to low room temperature, January 23, 2025 (5)

'D' EDG due to elevated water content in lube oil, January 31, 2025

(6) Reactor core isolation cooling due to elevated temperatures on the test return line to the condensate storage tank, March 13, 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) Trip and auto-start logic of safety-related heating, ventilation, and air conditioning trains, Design Change Package 80128401, March 21, 2025

71111.24 - Testing and Maintenance of Equipment Important to Risk

The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:

Post-Maintenance Testing (PMT) (IP Section 03.01) (6 Samples)

(1) Post-maintenance comprehensive testing following internal inspection of 'A' standby liquid control pump fluid end, January 7, 2025 (2)

'D' EDG lube oil heat exchanger plugging and lube oil replacement, February 7, 2025

(3) Repairs following the 'F' filtration, recirculation, and ventilation system failure to achieve desired flow, February 11, 2025 (4)

'C' refuel floor exhaust radiation monitor power supply replacement, February 17, 2025 (5)

'A' control room chiller breaker replacement, March 6, 2025 (6)

'B' control room emergency filtration system functional test following inspections of motor control center starter control units, March 24, 2025

Surveillance Testing (IP Section 03.01) (2 Samples)

(1) Reactor core isolation cooling high turbine exhaust pressure calibration and testing, January 16, 2025

(2) Class 1E 4.16kV feeder degraded voltage instrumentation channel functional test, February 10, 2025

Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) HC.OP-IS.BC-0001, "AP202, 'A' Residual Heat Removal Pump In-Service Test," January 2, 2025

71114.02 - Alert and Notification System Testing

Inspection Review (IP Section 02.01-02.04) (1 Sample)

(1) The inspectors evaluated PSEG's maintenance and testing of the station alert and notification system on March 17 through March 20, 2025, for the period of June 2023 through February 2025.

71114.03 - Emergency Response Organization Staffing and Augmentation System

Inspection Review (IP Section 02.01-02.02) (1 Sample)

(1) The inspectors evaluated the readiness of PSEG's Emergency Preparedness Organization on March 17 through March 20, 2025.

71114.04 - Emergency Action Level and Emergency Plan Changes

Inspection Review (IP Section 02.01-02.03) (1 Sample)

(1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes:

• 2024-09, E-plan, Section 1, "Introduction," Revision 18

• 2024-15, E-plan, Section 15, "Exercises and Drills," Revision 18

• 2024-18, E-plan, Section 6, "Notification Methods - Response Organizations,"

Revision 22

• 2024-19, E-plan, Section 2, "Assignment of Responsibility," Revision 26

• 2024-36, E-plan, Section 7, "Emergency Communications," Revision 19

• 2024-37, E-plan, Section 11, "Protective Response," Revision 17

• 2024-39, EP-HC-325-236, Attachment 6, "Hope Creek Emergency Action Level Radiological Setpoint Calculation Document," Revision 2

• 2024-41, EP-HC-325-103, "Offsite Rad Conditions," Revision 3; EP-HC-325-140, "HC EAL Wall Chart (All Conditions)," Revision 2; EP-HC-325-203, "Offsite Rad Conditions," Revision 2, Hope Creek Radiological Emergency Action Level Changes

This evaluation does not constitute NRC approval.

71114.05 - Maintenance of Emergency Preparedness

Inspection Review (IP Section 02.01 - 02.11) (1 Sample)

(1) The inspectors evaluated PSEG's maintenance and testing of the emergency preparedness program on March 17 through March 20, 2025, for the period of June 2023 through February 2025.

71114.06 - Drill Evaluation

Additional Drill and/or Training Evolution (1 Sample)

The inspectors evaluated:

(1) Licensed operators during simulator training evolutions, with associated emergency preparedness drill and exercise performance criteria on January 27,

71124.06 - Radioactive Gaseous and Liquid Effluent Treatment

Walkdowns and Observations (IP Section 03.01) (4 Samples)

The inspectors evaluated the following radioactive effluent systems during walkdowns:

(1) South plant ventilation system

(2) Turbine building circulating water system

(3) Liquid radwaste release system

(4) Cooling tower blowdown to the river

Sampling and Analysis (IP Section 03.02) (4 Samples)

Inspectors evaluated the following effluent samples, sampling processes and compensatory samples:

(1) OAT-303, Liquid Waste Floor Drain Sampling Tank Sample and Analysis

(2) OBT-346, Liquid Waste Component Leakage Collection Tank Sampling and Analysis

(3) North Plant Ventilation Weekly Sampling for Tritium, Particulate, Iodine, and Noble Gas (also was a 12-hour sample for the Noble Gas out of service)

(4) South Plant Ventilation Weekly Sampling for Tritium, Particulate, Iodine, and Noble Gas

Dose Calculations (IP Section 03.03) (2 Samples)

The inspectors evaluated the following dose calculations:

(1) G-20241201-404-C, Filtration Recirculation Ventilation System Continuous Release

(2) L-20230221-247-B, Waste Sampling Tank 'A' Batch Release

Abnormal Discharges (IP Section 03.04) (1 Sample)

The inspectors evaluated the following abnormal discharges:

(1) Tritium identified in the circulating water system on April 9, 2024, noted in Notification (NOTF) 20962541

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &

Transportation

Radioactive Material Storage (IP Section 03.01)

The inspectors evaluated the licensees performance in controlling, labeling and securing the following radioactive materials:

(1) Low level radwaste building boxes and liners

(2) North yard area boxes, trailers, and Sealand containers

Radioactive Waste System Walkdown (IP Section 03.02) (2 Samples)

The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:

(1) Equipment drains water processing

(2) Floor drain water processing

Waste Characterization and Classification (IP Section 03.03) (2 Samples)

The inspectors evaluated the following characterization and classification of radioactive waste:

(1) Dry active waste in a 20 foot Sealand container #298815

(2) Dry active waste in a 20 foot Sealand container #298855

Shipment Preparation (IP Section 03.04) (1 Sample)

(1) The inspectors observed the preparation of radioactive shipment HC25-008 of dry active waste on February 5, 2025.

Shipping Records (IP Section 03.05) (4 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) Shipment #HC24-001 Irradiated components, Class B, Cat 1

(2) Shipment #HC24-120 Irradiated components, Class B, Cat 1

(3) Shipment #HC23-095 Irradiated components, Class B, Cat 1

(4) Shipment #HC23-060 Resin Liner, Class A

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===

(1) January 1, 2024 through December 31, 2024

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (1 Sample)

(1) January 1, 2024 through December 31, 2024

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)

(1) January 1, 2024 through December 31, 2024

EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)

(1) April 1, 2024 through December 31, 2024

EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)

(1) April 1, 2024 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). PSEG 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. Verifications will take place during later inspections.

71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) Root cause evaluation, WO 70235435, of an as-found May 2024 ILRT failure, March 31, 2025

(2) Review PSEG's evaluation and corrective actions, WO 70236844, associated with the August 9, 2024, loss of 'C' 4kV vital bus

71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Follow-Up (IP Section 03.01)

(1) The inspectors evaluated 'D' EDG repetitive inoperability and PSEG's response on February 24, 2025.

Event Report (IP Section 03.02) (1 Sample)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000354/2024-001-00, 01 "Invalid Primary Containment Integrated Leak Rate As-Found Test (ILRT) Supplement," (ML24184C067), and updated LER submittal 05000354/2024-001-01 (ML24253A265). The inspection conclusions associated with these LERs are documented in this report under Inspection Results section 71152.

These LERs are closed.

INSPECTION RESULTS

Inadequate Procedure for Calibration of the Voltage and Current Transducers Associated with a 4 Kilovolt (kV) Vital Bus Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events

Green NCV 05000354/2025001-01 Open/Closed

None (NPP)71152A A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified when PSEG failed to maintain a calibration procedure for a safety-related 4kV vital bus.

Description:

On August 9, 2024, during preventive maintenance on the bus's voltage and current transducers, AT-6352A2, maintenance personnel were performing step 5.1.5.C of procedure HC.IC-DC.ZZ-0143, "Scientific Columbus Voltage and Current," Revision 12, which had technicians insert a non-conductive material between the shorting links of the test switch TS-06C-40308. Completion of this step resulted in the loss of power to the 'C' 4kV vital bus and entry into TS Limiting Condition for Operation (LCO) 3.8.3.1. The technical aspects of the loss of the C 4kV vital bus are discussed in NRC integrated inspection report 05000354/2024003 (ML24297A057).

The inspectors reviewed the calibration procedure, HC.IC.DC.ZZ-0143, personnel statements from the initiating event, and material used during the calibration. The inspectors noted that HC.IC-DC.ZZ-0143 does not specify the type of material, rigidity, size associated with non-conductive material, nor instructions of how the material should be inserted to isolate the current transformer (CT) circuit. The inspectors also noted that the rubber material used during the calibration was not consistent with historical practice; technicians that frequently perform these calibrations are trained to use a popsicle stick or piece of lamacoid.

The inspector reviewed PSEG's Apparent Cause Evaluation (ACE) 70236844 and noted that the procedure for the transducer calibration, HC.IC-DC.ZZ-0143, is a generic procedure for the calibration of Scientific Columbus Voltage and Current Transducers, Models VT110PA7 and CT510PA7. The ACE documented that adequate procedural guidance did not exist for use of a non-conducting piece of material. As stated in the ACE, "the act of inserting the material agitated the test switch blades, which caused a momentary interruption of current on the 51N relay. The 51N relay detects a neutral fault current by measuring a sum of currents from the 3-phases (3-CTs). With a disruption of current caused by the TS-06C-40308, the resulting change in the sum of the currents triggered the 51N and subsequently caused actuation of the lockout relaying."

The inspectors also noted that PSEG performed an extent of condition review and identified eight other preventive maintenance tasks that could produce a similar condition or consequence. Work order (WO) 70236844 documented a corrective action to place administrative holds on active work orders and revise procedures to clarify use of a "non-conducting piece of material."

Corrective Actions: Undervoltage lock out relay 860C1-403 was reset and the 'C' vital bus was re-energized. Similar work orders were placed on administrative hold pending procedure revisions.

Corrective Action References: NOTFs 20972856 and WO 70236844

Performance Assessment:

Performance Deficiency: The inspectors determined that PSEG's failure to maintain the calibration procedure for a 4kV vital bus was a performance deficiency. Specifically, PSEG's procedure HC.IC-DC.ZZ-0143 did not provide adequate instruction for the use of a non-conducting piece of material which resulted in the loss of a 4kV vital bus.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, PSEGs failure to maintain the calibration procedure resulted in the loss of a 4kV vital bus.

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 the finding required a detailed risk evaluation because the condition resulted in a partial loss of a support system (e.g., AC power). Specifically, a 4kV vital bus, 10A403, was de-energized resulting in the loss of power distribution Channel 'C'.

A Region I senior reactor analyst (SRA) performed the detailed risk evaluation and estimated the increase in core damage frequency associated with this performance deficiency to be 5.47E-8/yr, or of very low safety significance, Green. This was associated with an estimated increase in internal event risk given the loss of the 10A403 4kV vital bus.

Background

The trip of the 4kV vital bus 10A403 resulted in the loss of various equipment including, but not limited to:

• A control rod drive pump tripped

• C station service water pump tripped

• A primary containment instrument gas (PCIG) compressor tripped

The SRA noted that the associated C safety auxiliaries cooling system (SACS) pump was not in service prior to the loss of the bus, and therefore, SACS cooling to equipment loads was not affected by the scenario. The B SACS loop had two pumps in service and was supplying cooling to the turbine auxiliary cooling system loads. Therefore, there was no loss of cooling to the turbine auxiliary cooling system, which is the system that cools the instrument air compressor, secondary condensate pumps, and reactor high-pressure feedwater pumps. The SRA noted that the A PCIG compressor was in service and was lost during the event. This supplies the actuators of the inboard main steam isolation valves and maintains them in the open position at-power. The SRA noted that the backup B PCIG compressor is designed to start on lowering pressure. The operator placed the compressor in auto-lead from the auto-position to enable it to start in response to the loss of the A PCIG compressor. This action would prevent the inboard main steam isolation valves from eventually drifting closed on loss of gas pressure. The A station service water pump auto-started in place of the C station service water pump and SACS temperatures remained stable. Lastly, the operators started the B control rod drive pump to support the control rod drive system function which included maintaining the control rod scram accumulators and charging water pressure.

The SRA noted that the biggest risk impact of the loss of 4kV bus event, (frequency of 1.0/yr)compared to its baseline frequency of 4E-3/yr, was the potential challenge in preventing a loss of the power conversion system (PCS) and maintaining the inboard main steam isolation valves open. Therefore, the SRA modified the Hope Creek Standardized Plant Analysis Risk (SPAR) model to inform the risk of the event.

SPAR Model Information and Modifications The SRA developed the risk estimate for the failure to maintain 4kV bus 10A403 energized using System Analysis Program for Hands-On Integrated Reliability Evaluations (SAPHIRE),

Version 8.2.11, SPAR model Version 8.81, for Hope Creek. SPAR model changes and insights to reflect the current nominal as-built, as-operated unit as well as this specific condition included the following:

• The existing initiating event, IE-LOACB-B, was used as a surrogate for the loss of the C 4kV bus. The flag set was revised to include basic events, ACP-BAC-LP-10A403 and new event, HE-LOACB-403, both set to TRUE.

• The PCS fault tree was modified to include a leg to serve as a surrogate for the increased probability of losing the inboard main steam isolation valves by using three basic events, (HE-LOACB-403, IGS-MDP-FS-BK202, and IGS-XHE-FO-V5125) in an AND gate. The failure to align instrument air was set to ignore to provide a bounding assessment. IGS-MDP-FS-BK202 represented the failure of the B PCIG compressor to start.

• Numerous other changes were made to the model to reflect the current as-built plant configuration such as removal of the gas turbine generator and addition of the allowed outage time diesel generator.

A change set was created to simulate the loss of AC bus event:

• ADS-XHE-XM-NONADS, was set to TRUE.

• IE-LOACB-B (existing event in model) was set to 1.0/yr to simulate the loss of the C 4kV bus 10A403.

The cutsets given the loss of the 10A403 4kV bus were then sliced to reflect the conditional increase in the loss of the PCS given a failure of the B PCIG compressor to start.

Specifically, the plant transient or PCS risk increase given the loss of bus event was assessed through the review of the conditional cutsets. The dominant core damage sequence was the loss of the 4kV 10A403 bus, failure of high-pressure injection, and failure of the B PCIG compressor to start (surrogate for PCS event).

The potential increase in an actual plant transient due to the loss of the bus was assumed to be the dominant contributor to the increase in core damage frequency. The SRA noted that the bus was recovered in a nominal five hours. The SRA determined that the loss of mitigating systems risk was bounded by the potential initiating event plant transient risk increase.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance. Procedure HC.IC-DC.ZZ-0143, Revision 12, was last updated on November 27, 2012.

Enforcement:

Violation: Technical Specification 6.8.1 requires, in part, that written procedures shall be established, implemented, and maintained covering the activities referenced in the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Section 9 of Appendix A of Regulatory Guide 1.33, Revision 2, requires in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this, prior to August 9, 2024, PSEG's procedure HC.IC-DC.ZZ-0143 was not maintained to ensure adequate instruction for use of a non-conducting piece of material which resulted in the loss of a 4kV vital bus.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Inadequate Airlock Installation Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity

Green NCV 05000354/2025001-02 Open/Closed

None (NPP)71152A A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when PSEG did not provide appropriate quantitative or qualitative acceptance criteria in a containment personnel airlock installation procedure to determine it had been satisfactorily accomplished.

Description:

The Hope Creek containment and its penetrations are leak rate tested as required by 10 CFR 50, Appendix J. Appendix J, in part, requires the performance of Type A and Type B tests for primary containment leakage. Type A tests measure the containment overall integrated leak rate and are preceded by pressurization and stabilization evolutions.

Type B tests measure leakage across pressure-containing or leakage-limiting containment penetrations and boundaries other than containment isolation valves.

The Hope Creek drywell has two equipment hatches to allow for the removal of larger components for maintenance. One of these, the C2 equipment hatch, is normally fitted with a removable personnel airlock to allow for personnel access to primary containment. The airlock is bolted to the drywell equipment hatch via a flange and sealed using a dual O-ring design between the two seating surfaces. The airlock can be local leak rate tested and is considered a Type B test component.

On May 2, 2024, PSEG commenced a Type A containment ILRT. During containment pressurization, PSEG identified an audible, circumferential leak where the personnel airlock is bolted to the drywell equipment hatch. Since this leakage prevented satisfaction of the ILRT stabilization criteria, PSEG applied a pressure block between the C2 equipment hatch O-rings to identify and stabilize the leakage rate. On May 3, 2024, after not seeing improvement of the leakage rate, PSEG directed the application of a second pressure rig to the pressure block. Approximately one hour after the second pressure block was applied, operators observed a step-change decrease in the leakage rate indicative that the outer O-ring had changed the leakage flow. PSEG technicians also identified a visible 1/16 inch gap between a portion of the flange and the hatch while the remainder of the interface had metal-to-metal contact. PSEG classified the O-ring shift as an adjustment and declared the as-found ILRT a failure. PSEG then tightened the hatch bolts around the identified leak until the visible gap was closed. This allowed the ILRT stabilization criteria to be met and PSEG completed the Type A test.

PSEG reported the event to the NRC as a degraded condition in accordance with 10 CFR 50.72(b)(3)(ii)(A) in Event Notification 57103 on May 3, 2024. Subsequently, PSEG submitted LER 2024-001-00 to the NRC on July 2, 2024, per 10 CFR 50.73(a)(2)(ii)(A), for a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

Following the completion of a root cause evaluation under work order (WO) 70235435, PSEG submitted LER supplement 2024-001-01 and identified an additional criterion under 10 CFR 50.73(a)(2)(v)(C) for a condition that could have prevented fulfillment of a safety function of systems needed to control the release of radioactive material.

Inspectors reviewed PSEGs root cause evaluation and noted it identified inadequate procedural guidance for validating and measuring flange to drywell airlock surface metal-to-metal contact and absence of steps to validate torque values after rigging is relaxed or removed. In October 2007, PSEG removed and reinstalled the C2 airlock for inspections using procedure HC.MD-CM.XX-0002, Personnel Airlock/Equipment Hatch Removal and Replacement, Revision 10. This procedure directed that all torquing of the bolts be done while the hatch was supported by rigging. There were no steps to verify proper torquing of the bolts after the removal of rigging or to verify metal-to-metal contact between the flange and the hatch. Additionally, design drawings showed the mating surface outside of the O-rings as a metal-to-metal interface. Despite the procedural inadequacies, PSEG completed a satisfactory local leak rate test following airlock reinstallation in November 2007. From that time until the ILRT in 2024, a satisfactory ILRT and two satisfactory local leak rate tests had been performed with leakage below administrative limits. No maintenance had been performed on the C2 drywell hatch flange from 2007 until the ILRT in 2024.

Corrective Actions: PSEG re-torqued the C2 hatchs bolts and performed a satisfactory as-left Type A ILRT to verify that primary containment integrity existed. PSEG also revised procedure HC.MD-CM.XX-0002 to ensure metal-to-metal surface contact between the drywell flange and the airlock after the first high-pressure torque pass and to perform additional high-pressure torque passes after the rigging is relaxed until there is no additional nut movement.

Based on the as-found failure, PSEG removed the C2 drywell hatch flange local leak rate test from its extended nine-year frequency and returned it to a 30-month frequency.

Corrective Action References: NOTF 20960149, WO 70235435

Performance Assessment:

Performance Deficiency: The inspectors determined that having an inadequate reinstallation procedure for the personnel airlock in accordance with 10 CFR 50, Appendix B, Criterion V, was a performance deficiency. Specifically, PSEG procedure HC.MD-CM.XX-0002, "Personnel Airlock/Equipment Hatch Removal and Replacement," did not verify proper torquing of the bolts after the removal of rigging or metal-to-metal contact between the flange and the hatch, as required by design.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the issue resulted in a principal safety barrier being seriously degraded and could have prevented the fulfillment of the safety function to control the release of radioactive material. The inspectors also noted that this issue was similar to IMC 0612, Appendix E, Example 4.m, because PSEGs procedure was inadequate to ensure that a quality component was properly reinstalled.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors started the finding significance assessment using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process given that discovery occurred in a refueling outage. The inspectors used Appendix G, Attachment 1, Exhibit 4, and determined that the finding should be assessed using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, because the finding degraded the physical integrity of the reactor containment. Using IMC 0609, Appendix H, inspectors determined this was a Type B finding since it was related to a degraded containment integrity condition without affecting core damage likelihood and with a likely duration of greater than 30 days. Inspectors then used IMC 0609, Appendix H, Table 7.3 for shutdown findings, where Note 2 directed that the finding should be assessed as a Type B finding at-power, because the finding may have existed before shutdown and could have impacted large early release frequency upon change of containment status given a return to power. The finding did not screen out using Table 7.1 for a boiling water reactor Mark I containment at-power and required a phase 2 assessment. The inspectors determined that the finding screened to Green in the phase 2, using IMC 0609, Appendix H, Table 7.2, because the leakage from the drywell was less than 100 percent containment volume/day. For thoroughness, the inspectors also used IMC 0609, Appendix A, Exhibit 3, for significance of Barrier Integrity findings at-power which also directed use of IMC 0609, Appendix H, when the finding represents an actual open pathway in containment physical integrity.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance. The airlock installation procedure was last implemented in 2007.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, from October 15, 2007 to May 3, 2024, PSEGs procedure for airlock installation did not provide appropriate acceptance criteria to determine that the activity had been satisfactorily accomplished.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Observation: Review of Corrective Actions and Extent of Condition for Loss of 'C' 4kV Vital Bus 71152A From February 3 through 7, 2025, inspectors reviewed PSEG's ACE 70236844 and associated corrective actions for the loss of the 'C' 4kV vital bus during testing on August 9, 2024. The inspectors identified one non-cited violation that is documented in the Inspection Results section. Additionally, inspectors reviewed three corrective actions generated to address guidance in procedures HC.IC-DC.ZZ-0143, HC.IC-DC.ZZ-0150, and HC.IC-DC.ZZ-0155 that are used to perform 4kV vital bus testing. These procedure changes were due on January 31, 2025, but were incomplete at the time of the inspection. PSEG procedure LS-AA-125, "Corrective Action Program," Revision 29, requires that completion of corrective actions for conditions adverse to quality be tracked in the corrective action program. Contrary to LS-AA-125, the inspectors identified that these corrective actions were closed to work orders in a non-corrective action program without the procedure revisions being completed. This improper closure of corrective actions was a minor performance deficiency because PSEG had placed preventive maintenance work orders associated with these procedures on administrative hold until the procedures were revised. PSEG documented this concern in NOTF 20986605.

Unresolved Item (Open)

Unresolved Item Regarding the NRC's Grant of a Notice of Enforcement Discretion (NOED) for an EDG under TS 3.8.1.1 URI 05000354/2025001-03 71153

Description:

On January 27, 2025, at 12:00 a.m., Hope Creek declared the A EDG inoperable and removed it from service for planned maintenance. On January 31, 2025, at 7:35 a.m., the D EDG was declared inoperable due to an emergent lube oil system issue.

Hope Creek entered TS LCOs 3.8.1.1.e and 3.8.1.1.f for two inoperable EDGs. Action 3.8.1.1.f was met; however, Action 3.8.1.1.e required restoration of one inoperable EDG to operable status within two hours (9:35 a.m.) or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (9:35 p.m.) and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (February 1, 2025, at 9:35 p.m.). PSEG determined that the A EDG work was progressing as planned and was expected to be complete no later than February 2, 2025, at 1:00 a.m. By letter (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25034A230) dated February 3, 2025, PSEG requested the NRC exercise discretion to not enforce compliance with the actions required by Hope Creek TS LCO 3.8.1.1.e and extend the expiration of the completion time by 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> to February 2, 2025, at 1:35 a.m., to allow Hope Creek additional time to return the A EDG to operable status.

Based on the NRC staffs evaluation of PSEGs request, the NRC determined that granting this NOED was consistent with the NRCs Enforcement Policy and staff guidance and the request met the criteria specified in NRCs Enforcement Manual, Appendix F. As communicated verbally to PSEG at approximately 9:15 p.m., on January 31, 2025, the NRC exercised discretion to not enforce compliance with TS Action LCO 3.8.1.1.e for the 28-hour period from January 31, 2025, at 9:35 p.m., until February 2, 2025, at 01:35 a.m. Hope Creek returned the A EDG to operable status on January 31, 2025, at 9:21 p.m., and the approved NOED was not utilized. The NRC issued a letter to PSEG documenting the grant of this NOED on February 5, 2025 (ML25036A046).

While the NOED was not utilized, enforcement action may be taken to the extent that violations were involved for the root cause that led to the noncompliance for which the NOED was necessary. In accordance with IMC 0611, 06.05.o, when a NOED is granted, an unresolved item shall be opened in accordance with the NRC Enforcement Manual, Appendix F, Notices of Enforcement Discretion, to track the required follow-up activities to determine if there was a performance deficiency or if additional information is required to determine if the NOED issue was more than minor.

Planned Closure Actions: NRC follow-up activities shall be conducted in accordance with Inspection Procedure 71153, Section 03.04, and documented in accordance with IMC 0611.

Licensee Actions: Hope Creek returned the A EDG to operable status on January 31, 2025, at 9:21 p.m., and exited TS Action LCO 3.8.1.1.e.

Corrective Action References: NOTF 20986945. WO

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On April 9, 2025, the inspectors presented the integrated inspection results to Eric Larson, Site Vice President, and other members of the licensee staff.

• On February 6, 2025, the inspectors presented the 71124.08 inspection results to Eric Larson, Site Vice President, and other members of the licensee staff.

• On March 20, 2025, the inspectors presented the emergency preparedness program inspection results to Eric Larson, Site Vice President, and other members of the licensee staff.

• On March 20, 2025, the inspectors presented the 71124.06 inspection results to Eric Larson, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.04

Corrective Action

Documents

Resulting from

Inspection

20987236

20987237

20987447

71111.05

Corrective Action

Documents

Resulting from

Inspection

20985912

20986582

20987436

20987934

20987310

20987861

20987875

20973991

20990580

20988895

20990611

20988909

71111.11Q Corrective Action

Documents

Resulting from

Inspection

20986580

71111.12

Corrective Action

Documents

Resulting from

Inspection

20987379

71111.13

Corrective Action

Documents

Resulting from

Inspection

20986563

20986361

20987310

20988652

71111.15

Corrective Action

Documents

Resulting from

Inspection

20986279

20986356

20985583

20988140

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Miscellaneous

NOTF 20985736

71111.18

Corrective Action

Documents

Resulting from

Inspection

20988857

71111.24

Corrective Action

Documents

Resulting from

Inspection

20985015

20985401

20985510

20986700

20988769

20988856

Miscellaneous

NOTF 20985280

71114.02

Miscellaneous

Final 2005 REP-10 Design Review Report, PSEG Salem

and Hope Creek Generating Stations

Revision 1

71114.03

Miscellaneous

Hope Creek Generating Station and Salem Generating

Station On-Shift Staffing Analysis Report

Revision 0

Procedures

EP-AA-120-1007

Maintenance of Emergency Response Organization

Revision 11

71114.04

Procedures

EP-AA-120-1001

CFR 50.54(q) Change Evaluation

Revision 5

71114.05

Miscellaneous

PSEG Nuclear LLC Emergency Plan

71124.08

Corrective Action

Documents

Resulting from

Inspection

20981593

Identified Leakage from Flanges for Chemical Addition Pump

0AP-333

2/04/2025

20986635

Room 3101 Lights Out

2/04/2025

71152A

Corrective Action

Documents

Resulting from

Inspection

20986007

20986205

20987000

20987409

20990693

20986752

20986606

20986605

Drawings

E-0046-1

Schematic Meter & Relay Diagram 4.16 KV IE Station Power

System Switchgears 10A401 & 10A403

Revision 10

Procedures

LS-AA-120

Issue Identification and Screening Process

Revision 25

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Work Orders

279913

9Y PM CAL 10-A-403 4160 SWGR CH 'C' IND