IR 05000354/2023003

Integrated Inspection Report 05000354/2023003

ML23311A046
Person / Time
Site:	Hope Creek
Issue date:	11/07/2023
From:	Brice Bickett Division of Operating Reactors
To:	Mcfeaters C Public Service Enterprise Group
References
IR 2023003
Download: ML23311A046 (1)

Text

November 7, 2023

SUBJECT:

HOPE CREEK GENERATING STATION - INTEGRATED INSPECTION REPORT 05000354/2023003

Dear Charles McFeaters:

On September 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Hope Creek Generating Station. On October 18, 2023, the NRC inspectors discussed the results of this inspection with Robert DeNight, 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. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

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

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Hope Creek Generating Station.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; 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, Brice A. Bickett, 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:

05000354/2023003

Licensee:

PSEG Nuclear, LLC

Facility:

Hope Creek Generating Station

Location:

Hancocks Bridge, NJ

Inspection Dates:

July 01, 2023 to September 30, 2023

Inspectors:

J. Patel, Senior Resident Inspector

J. Bresson, Resident Inspector

B. Edwards, Health Physicist

E. Eve, Senior Project Engineer

A. Kostick, Health Physicist

J. Kulp, Senior Reactor Inspector

Approved By:

Brice A. Bickett, 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

Improper Assembly and Inadequate Post-Maintenance Testing of a Drywell to Suppression Chamber Vacuum Breaker Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000354/2023003-01 Open/Closed

[H.5] - Work Management 71152A A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when PSEG did not ensure the preventive maintenance (PM)procedure was adequate to provide instructions for reassembly and calibration of safety-related drywell to suppression chamber vacuum breakers. Consequently, the 'A' vacuum breaker was improperly reassembled, with mechanical interference between the vacuum breaker pallet (closure obturator) and the limit switches and closing magnets that prevented the pallet from fully seating. This led to a bypass leak between the drywell and the suppression chamber and challenged the primary containment integrity.

Failure to Identify and Correct a Condition Adverse to Quality (CAQ) Related to Suppression Chamber Pressures Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000354/2023003-02 Open/Closed

[H.12] - Avoid Complacency 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for PSEGs failure to promptly identify and correct a condition adverse to quality. Specifically, the inspectors identified that PSEG screened Notification (NOTF) 20919478 which documented an adverse trend associated with drywell to suppression chamber pressures during normal nitrogen makeup as significance level 4, or a condition not classified as CAQ, and therefore a degraded condition. As a result, the issue was not evaluated and corrected in a timely manner and called into question the operability of primary containment.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000354/2023-001-00 LER 2023-001-00 for Hope Creek Generating Station,

Technical Specification Required Shutdown Due to 71153 Closed

Declaring the Suppression Chamber and Primary Containment Inoperable

LER 05000354/2023001-01 LER 2023-001-01 for Hope Creek Generating Station,

Technical Specification Required Shutdown due to Declaring the Suppression Chamber and Primary Containment Inoperable 71153 Closed LER 05000354/2023-001-02 LER 2023-001-02 for Hope Creek Generating Station,

Technical Specification Required Shutdown due to Declaring the Suppression Chamber and Primary Containment Inoperable 71153 Closed

PLANT STATUS

The Hope Creek Generating Station began the inspection period at rated thermal power. On July 27, 2023, the unit was downpowered to 90 percent due to the main condenser pressure limit affected by atmospheric conditions and returned to rated thermal power on July 30, 2023. On September 1, 2023, the unit was downpowered to 87 percent to perform control rod pattern adjustment and returned to rated thermal power on September 2, 2023. On September 16, 2023, the unit was downpowered to 69 percent to perform turbine valve testing and control rod pattern adjustment. The unit returned to rated thermal power on September 17, 2023, and 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.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (3 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Service water system, July 19, 2023 (2)

'B' primary containment instrument gas system, August 22, 2023 (3)

'B' residual heat removal (RHR), September 27, 2023

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the

'A' emergency diesel generator (EDG) and associated 4.16kV bus during 'B' EDG 24-hour endurance test the week of September 11, 2023.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 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) Circulating water structure in pre-fire plan FP-HC-3712, July 24, 2023

(2) Control area heating, ventilation, and air conditioning equipment rooms in pre-fire plan FP-HC-3563, August 1, 2023

(3) High pressure coolant injection pump and turbine room, RHR pump room, and heat exchanger rooms in pre-fire plan FP-HC-3413, August 16, 2023

(4) RHR pump room, RHR heat exchanger room, and electrical equipment room in pre-fire plan FP-HC-3412, August 16, 2023

(5) Heating, ventilation, and air conditioning equipment rooms in pre-fire plan FP-HC-3571, September 25, 2023 (6)

'B' RHR pump, heat exchanger, and valves rooms in pre-fire plans FP-HC-3412 and FP-HC-3422, September 27, 2023

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1)1A2E201 safety auxiliaries cooling system heat exchanger clean and inspect during refueling outage H1R24, July 27, 2023

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 a downpower for turbine valve testing and control rod pattern adjustment on September 16, 2023.

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

(1) The inspectors observed and evaluated licensed operator performance in the simulator during an out-of-the-box training exercise on July 11, 2023.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (3 Samples)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1)

'B' control area ventilation system following an automatic trip, on July 4, 2023, due to low flow conditions, August 10, 2023

(2) Scram discharge volume drain and vent valves following failure to close, August 25, 2023 (3)

'A' EDG following a speed switch failure, week of September 25, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1)

'C' EDG planned inoperability, week of July 10, 2023

(2) Emergent inoperability of 'A' EDG, August 2-3, 2023

(3) Planned maintenance activities (high pressure coolant injection limiting condition for operation (LCO), 'A' control room emergency filtration system LCO, and 'C' RHR pump LCO), week of September 5, 2023

(4) Unplanned inoperability of 'B' EDG, week of September 11, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)

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

(1) Safety and turbine auxiliary cooling system supply and return isolation valves leak-by, July 13, 2023

(2) Control rod 54-35 not indicating fully withdrawn, August 4, 2023 (3)

'A' EDG after the jacket water pressure switch not initiating a generator field flash, August 8, 2023 (4)

'B' RHR pump following indication of oil leak from the motor oil seal, September 28, 2023

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)

'A' EDG speed switch replacement with a newer version, September 6, 2023

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) (7 Samples)

(1)

'B' service water pump replacement, July 26, 2023 (2)

'A' EDG speed switch replacement, August 3, 2023 (3)

'B' EDG output breaker following remote operation functional test, August 14, 2023

(4) Reactor manual control system power supply and transponder card replacement, August 16, 2023 (5)

'D' RHR pressure transmitter trip unit replacement, August 23, 2023 (6)

'C' RHR pump planned maintenance, week of September 4, 2023 (7)

'B' EDG fuel injector line repair, September 12, 2023

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) HC.OP-ST.KJ-0002, "'B' Emergency Diesel Generator Monthly Operability Run,"

July 17, 2023

(2) HC.OP-ST.GU-0005, "FRVS Operability Test (Single Recirculation Fan Method),"

August 15, 2023

(3) HC.OP-ST.KJ-0015, "EDG 1BG400 - 24 Hour Operability Run and Hot Restart Test,"

September 21, 2023

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

(1) HC.OP-IS.BJ-0101(Q), "High Pressure Coolant Injection System Valves Inservice Test," July 13,

71124.05 - Radiation Monitoring Instrumentation

Walkdowns and Observations (IP Section 03.01) (8 Samples)

The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:

(1)

"FRVSV SKID" Area Radiation Monitor (ARM) #6621 (2)

"FRVSV PROCESSOR" ARM #6621

(3) Aerial Monitoring System (AMS) (located inside of radiological controlled area (RCA))

1. 13177

(4) AMS (located inside of RCA) #12918

(5) Eberline RM-14 contamination monitor (located inside of RCA) #8212

(6) ADM-616 (ARM located inside of RCA) #2R18

(7) Instrument Room SAM #335001

(8) Canberra Accuscan 2 whole body counting system

Calibration and Testing Program (IP Section 03.02) (13 Samples)

The inspectors evaluated the calibration and testing of the following radiation detection instruments:

(1) Control Room ARM HOSD-0SDRE-6609

(2) Ludlum 30-7 detector #LUD30-7006

(3) Ludlum 30-7 detector #LUD30-7003

(4) Ludlum 9-4 detector #LUD9H316350

(5) Ludlum 9-4 detector #LUD9H316394

(6) Mirion Telepole #TPOLEH6611-053

(7) Mirion Telepole #TPOLEH6613-026

(8) Ludlum 9-4 detector #LUD9S316366

(9) Mirion Teletector #TELEE3717

(10) Eberline E-530N #E530NS1444

(11) Eberline E-520 #E520E1872

(12) Ludlum 3000 #LUD3000S-001

(13) Ludlum 14C #LUD14C236861

Effluent Monitoring Calibration and Testing Program Sample (IP Sample 03.03) (2 Samples)

The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:

(1) Effluent Gas Monitor, "South Ventilation Stack" RMS #1SPRY-487513

(2) Effluent Gas Monitor, "Rx BLDG Ventilation System Exhaust" RMS #1SPRY-4826

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

MS07: High Pressure Injection Systems (IP Section 02.06)===

(1) July 1, 2022 through June 30, 2023

MS08: Heat Removal Systems (IP Section 02.07) (1 Sample)

(1) July 1, 2022 through June 30, 2023

MS09: Residual Heat Removal Systems (IP Section 02.08) (1 Sample)

(1) July 1, 2022 through June 30, 2023

MS10: Cooling Water Support Systems (IP Section 02.09) (1 Sample)

(1) July 1, 2022 through June 30, 2023

===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) Test conditions for drywell to suppression chamber leak-by test could not be established.

71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

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

(1) LER 05000354/2023-001-00, Technical Specification Required Shutdown Due to Declaring the Suppression Chamber and Primary Containment Inoperable (ML23173A087), and updated LER submittals 05000354/2023-001-01 (ML23254A226) and 05000354/2023-001-02 (ML23269A205). The inspection conclusions associated with these LERs are documented in this report under Inspection Results Section 71152. These LERs are closed.

INSPECTION RESULTS

Improper Assembly and Inadequate Post-Maintenance Testing of a Drywell to Suppression Chamber Vacuum Breaker Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems

Green NCV 05000354/2023003-01 Open/Closed

[H.5] - Work Management 71152A A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when PSEG did not ensure the preventive maintenance (PM)procedure was adequate to provide instructions for reassembly and calibration of safety-related drywell to suppression chamber vacuum breakers. Consequently, the 'A' vacuum breaker was improperly reassembled, with mechanical interference between the vacuum breaker pallet (closure obturator) and the limit switches and closing magnets that prevented the pallet from fully seating. This led to a bypass leak between the drywell and the suppression chamber and challenged the primary containment integrity.

Description:

Hope Creek Generating Station's (Hope Creek) Updated Final Safety Analysis Report (UFSAR), Section 6.2.1.1, Pressure Suppression Containment, describes that the containment provides a means to channel the flow from postulated pipe ruptures in the drywell to suppression pool. The drywell and suppression pool are connected by eight equally spaced vent pipes. A vacuum breaker assembly is located at the suppression chamber end of each vent line to limit differential pressure between the drywell and suppression chamber.

UFSAR, Section 6.2.1.1.5, discusses the steam bypass of the suppression pool and describes as follows: there are no flow paths that have been identified, other than an open vacuum relief valve (vacuum breaker) between the suppression chamber and the drywell, that could permit the bypassing of steam from the drywell directly to the suppression chamber free space. Redundant position switches on each vacuum relief valve and frequent testing of valve operability ensures that the possibility of an open relief valve coincident with an accident situation need not be considered.

On April 30, 2023, operators attempted to conduct surveillance test procedure HC.OP-ST.ZZ-0006, Drywell to Suppression Chamber Leak Rate Test, Revision 19; however, they could not establish the initial test conditions for drywell to suppression chamber differential pressure as required in Surveillance Requirement 4.6.2.1.f. As a result, the licensee declared the primary containment inoperable and placed the unit in Cold Shutdown as required by Technical Specification (TS) 3.6.1.1, "Primary Containment Integrity." Subsequent investigation by the licensee determined that the drywell to suppression chamber vacuum breaker A had significant bypass leakage due to improper seating. Specifically, the vacuum breakers closing magnets and bottom right close limit switch were hindering the vacuum breaker from properly seating. PSEGs prompt corrective actions involved repair of the A vacuum breaker and verifying operability by performing procedures HC.OP-ST.GS-0004, Suppression Chamber/Drywell Vacuum Breaker Operability Test, Revision 15, and HC.OP-ST.ZZ-0006, Drywell to Suppression Chamber Leak Rate Test, Revision 19. The licensee entered this issue into their corrective action program and completed a root cause evaluation (RCE), RCE 70229221.

The inspectors reviewed RCE 70229221 and noted the direct cause was the drywell to suppression chamber vacuum breaker A was not set appropriately to prevent mechanical interference from its associated limit switches and closing magnet assemblies following maintenance in October 2022. The inspectors also noted that HC.MD-CM.GS-0002, Drywell to Torus and Torus to Reactor Building Vacuum Relief Valve Overhaul, Revision 18 was identified as not including all required original equipment manufacturer (OEM) calibration, reassembly, and PMT requirements which led to the A vacuum breaker being reassembled with mechanical interference that kept its pallet from fully seating.

Specifically, mechanical interference caused by reinstallation of limit switches and closing magnets resulted in the vacuum breaker being held open approximately 0.06 inch. The inspectors reviewed HC.MD-CM.GS-0002 and noted it directs to bolt limit switches back to the same position. The inspectors noted that a new vacuum breaker seal was installed during the October 2022 maintenance and with a new seal installed, the seating characteristic of the vacuum breaker changed, and bolting in limit switches to the same position prior to seal replacement may not be the correct position. The inspectors also noted that subsequent maintenance resulted in A vacuum breaker closing magnets adjustments which resulted in additional mechanical interference. The inspectors noted that PSEG procedure HC.MD-ST.GS-0001, Torus to Drywell Vacuum Relief Valve Testing, Revision 15 is used to demonstrate the vacuum breaker opening setpoint and indicator operability. The inspectors reviewed the OEM vendor manual, PM150AQ-013, and noted that it directs a paper test to verify seat tightness after magnet setpoint adjustment. The inspectors noted that HC.MD-ST.GS-0001, does not direct a paper test to verify seat tightness after magnets and limit switches setpoint adjustment.

The inspectors reviewed PSEG procedure MA-AA-716-012, Post-Maintenance Testing, Revision 23, and noted it requires performing a drywell to suppression chamber bypass leakage test to demonstrate the suppression chamber operability following any maintenance activities that involve complete or partial disassembly, or replacement of components on these vacuum breakers. This licensee requirement is met by conducting procedure HC.OP-ST.ZZ-0006. The inspectors reviewed WO history for the A vacuum breakers and noted that HC.OP-ST.ZZ-0006 was not performed following the October 2022 maintenance. The inspectors noted that RCE 70229221 documented that HC.OP-ST.ZZ-0006 was not planned because the Operations Department did not perform an operations retest review of the work order associated with maintenance on the A vacuum breaker as required by PSEG procedure OP-HC-109-115-1002, Outage Preparation and Execution Implementation Guidelines, Revision 3.

The inspectors determined that PSEG did not properly reassemble drywell to suppression chamber vacuum breaker A and did not adequately perform PMT following the October 2022 maintenance.

As a result of this event, PSEG submitted LER 05000354/2023-001-00, Technical Specification Required Shutdown Due to Declaring the Suppression Chamber and Primary Containment Inoperable, (ML23173A087) and two subsequent updates (ML23254A226, ML23269A205).

Corrective Actions: PSEGs prompt corrective actions involved repair of the A vacuum breaker and verifying operability by performing procedures HC.OP-ST.GS-0004, Suppression Chamber/Drywell Vacuum Breaker Operability Test, Revision 15, and HC.OP-ST.ZZ-0006, Drywell to Suppression Chamber Leak Rate Test, Revision 19.

Corrective Action References: 70229221, 20935223, 20932336, 20935114, 20935324

Performance Assessment:

Performance Deficiency: The inspectors determined that PSEGs failure to properly reassemble the drywell to the suppression chamber vacuum breaker and failure to perform an adequate PMT was a performance deficiency within their ability to foresee and correct and should have been prevented.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined this performance deficiency also affected the Barrier Integrity cornerstone. Specifically, the omission of important instructions for calibration, reassembly, and testing from the OEM vendor manual within the maintenance procedure resulted in the improper reassembly of a safety-related vacuum breaker.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Based on the review of the General Electric Hitachi (GEH) analysis and discussion with regional senior reactor analysts, the inspectors determined this performance deficiency is best assessed by using Exhibit 2, Mitigating Systems Screening Questions, since the vacuum breaker degraded condition did not result in a failure of containment pressure control/heat removal functions and the risk significance determined by assessing the mitigating system function. Using Exhibit 2 the finding screened to Green, very low safety significance, because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, or did not involve a loss of a probabilistic risk assessment function for greater than TS allowed outage time. The inspectors, in collaboration with the Region 1 senior reactor analysts, determined it was very low safety significance based on the following.

In the event of a loss-of-coolant accident (LOCA), with an open vacuum breaker, steam flow from the containment drywell would not all be delivered to the suppression pool through the Mark I containment vent system, with some amount going through the suppression chamber air space, via the degraded A vacuum breaker, resulting in a steam bypass condition which could affect the LOCA containment response. PSEG engaged GEH to evaluate this degraded condition. GEH analyzed using the NRC approved computer code (SHEX) for the long-term containment analysis, in which a design basis accident (DBA) LOCA containment response analysis and steam line break with steam bypass containment analysis were performed for Hope Creek with the assumption that one vacuum breaker partially opened during the entire transient. This diverted flow was modeled as a bypass leakage assumption in SHEX calculation. For conservatism, the bypass area was doubled from the measured vacuum breaker opening area in the analysis. The results of this calculation were provided to PSEG in the GEH analysis HSE230046, Hope Creek Generating Station DBA-LOCA and Steam Line Break with Bypass Due to Open Wetwell to Drywell Vacuum Breaker Containment Analysis, Revision 0. PSEG documented their review of the GEH analysis in Technical Evaluation 70229221-0010 and concluded that primary containment pressure and temperature would remain within their corresponding design limits for a suppression chamber to drywell vacuum breaker constantly held open 0.060 inches during a DBA-LOCA or steam bypass event.

The inspectors reviewed GEH analysis and PSEG technical evaluation. Based on the results from the GEH analysis, the inspectors determined that the degraded condition did not result in a loss of the primary containment safety function, because with A vacuum breaker remaining opened, it would not have resulted in a significant increase in primary containment pressure and temperature. As demonstrated in the GEH analysis, the DBA-LOCA containment analysis shows an increase of 0.5 psi in the predicted peak DBA-LOCA drywell pressure due to a bypass through a vacuum breaker. Considering a large margin of approximately 11.4 psi in drywell pressure from the containment analysis, this 0.5 psi increase does not impact the primary containment integrity. The results from the steam bypass event also show that the predicted containment pressure and temperature are below their respective design limits and the pressure suppression/heat removal remained functional.

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee did not ensure operations retest review of the work order associated with maintenance on the A vacuum breaker as required by OP-HC-109-115-1002.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to the above, from October 15, 2022, to October 30, 2022, HC.MD-CM.GS-0002, Drywell to Torus and Torus to Reactor Building Vacuum Relief Valve Overhaul, Revision 18, was used to conduct maintenance on the A drywell to suppression chamber vacuum breaker when the procedure was inappropriate, because it did not include all required instructions for calibration, reassembly, and testing requirements from the OEM vendor document PM150AQ-013.

Additionally, PSEG procedure MA-AA-716-012, Post-Maintenance Testing, Revision 24, requires drywell to suppression chamber bypass leakage test as a final PMT. Contrary to this, on October 30, 2022, PSEG did not perform the required PMT and retest to demonstrate that the suppression chamber is operable following the maintenance of the vacuum breakers.

Consequently, the degraded condition of the A vacuum breaker pallet not fully seated went misdiagnosed from October 18, 2022 to April 30, 2023. This resulted in declaring the suppression chamber and primary containment inoperable on April 30, 2023, and a Technical Specification required shutdown.

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

Failure to Identify and Correct a Condition Adverse to Quality (CAQ) Related to Suppression Chamber Pressures Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems

Green NCV 05000354/2023003-02 Open/Closed

[H.12] - Avoid Complacency 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for PSEGs failure to promptly identify and correct a condition adverse to quality. Specifically, the inspectors identified that PSEG screened Notification (NOTF) 20919478 which documented an adverse trend associated with drywell to suppression chamber pressures during normal nitrogen makeup as significance level 4, or a condition not classified as CAQ, and therefore a degraded condition. As a result, the issue was not evaluated and corrected in a timely manner and called into question the operability of primary containment.

Description:

Hope Creek's TS LCO 3.6.2.1 requires that the suppression chamber shall be operable with a total leakage between the suppression chamber and drywell of less than the equivalent leakage through a 1-inch diameter orifice at a differential pressure of 0.80 psi. This LCO is met by implementing the TS Surveillance Requirement 4.6.2.1.f and conducting a drywell to suppression chamber bypass leakage test at an initial differential pressure of 0.80 psi and verifying that the differential pressure does not decrease by more than 0.24 inch of water per minute for a period of 10 minutes. Limiting the leakage from the drywell to the suppression chamber ensures the pressure suppression function of the primary containment is maintained. Thus, if an event were to occur that pressurized the drywell, the steam would be directed through downcomers into the suppression pool.

The inspectors reviewed NOTF 20919478, initiated on October 31, 2022, and noted that it documented an observation during the initial nitrogen make up to the drywell following inerting the primary containment. The operators in the main control room observed that the pressure in both the drywell and the suppression chamber increased simultaneously within 0.02 psi, while raising drywell pressure to 0.42 psi, at which point the nitrogen make up was secured. The NOTF further stated that the other main control room and control room integrated display system indicators displayed similar values, confirming the consistency of the data. Additionally, it was documented that, based on previous operating history, the drywell pressure exceeds the pressure in the suppression chamber during normal nitrogen makeup when raising the pressure to the nominal value of 0.7 psi, implying that there should not be any open pathway between drywell and suppression chamber that directly communicates with the air space in suppression chamber. The inspectors determined that the conditions described in NOTF 20919478 indicated potential bypass leakage from the drywell to the suppression chamber.

The inspectors reviewed licensee documents and interviewed operations staff and identified that the trend of drywell to suppression chamber pressure was consistently observed during daily nitrogen make ups to the drywell. The inspectors noted that NOTF 20919478 did not document that this trend occurred during each make up process. On April 20, 2023, the inspectors observed the normal daily nitrogen make up from the main control room and noted that both the drywell and suppression chamber pressures trended within 0.01 psi. The inspectors shared their observations with the licensee staff and raised concerns regarding potential bypass leakage from the drywell to the suppression chamber.

The inspectors noted that NOTF 20919478 was screened as a significance level 4 issue in accordance with the procedure LS-AA-120, Issue Identification and Screening Process, Revision 24. The inspectors reviewed LS-AA-120 and noted Sections 2.23.1 and 2.17.1 of this procedure, define significance level 4 as a non-corrective action program condition that does not meet the criteria for being a significant CAQ, a CAQ, or a condition affecting regulatory compliance. LS-AA-120, Attachment 2, Notification Level, provides specific examples and guidance to help determine the significance level. Step 4.4.2 in the procedure requires the Station Ownership Committee (SOC) to classify each issue with appropriate categorization in accordance with the Attachment 2. The SOC, responsible for screening each issue, classified NOTF 20919478 as significance level 4 on November 1, 2022.

Because of the non-corrective action program treatment of the NOTF, no corrective actions were assigned to evaluate and correct the deficiency. SOC's only assigned action was for engineering, an Action Tracking Item (ACIT) 70226369-0010, to review the trend data and initiate NOTF if required. In accordance with the procedure LS-AA-125, Corrective Action Program, Revision 29, an ACIT is assigned to minor problems that do not represent a CAQ, and is outside the corrective action program scope, and documentation is not required for extending the due date. The inspectors determined engineering did not take any actions under the assigned ACIT task to review and evaluate the trend data.

Additionally, the inspectors noted that NOTF 20919478 was updated on November 8, 2022, December 7, 2022, and February 10, 2023, which included additional information about the original condition. However, the notification was not returned to the SOC for a re-screening of the significance level and corrective action assignment, which are specifically required by LS-AA-120, step 4.2.5. Based on the additional information included in the NOTF, the inspectors concluded that the NOTF should have been returned to the SOC and should have been re-screened as significance level 2 (condition adverse to quality) in accordance with the 2 of LS-AA-125. As described in the Attachment 2, a condition adverse to quality includes any condition that has the potential to inhibit, or inhibited, a non-safety-related SSC from satisfactory performance of a function necessary for a safety-related SSC to fulfill its safety-related function. The inspectors found that the observed drywell to suppression chamber pressure trend during daily nitrogen make up in the primary containment represented a deficiency that could inhibit the primary containment from satisfactorily performing its safety function due to bypass leakage between the drywell and suppression chamber, thus challenging the integrity of the primary containment.

As a result of the inspectors concern, PSEG initiated NOTF 20934082 on April 19, 2023. In addition, the licensee rescreened NOTF 20919478 and assigned corrective actions including demonstrating primary containment operability by completing Technical Specification Surveillance Requirement 4.6.2.1.f. However, due to the inability to establish the necessary test conditions for the surveillance test, the suppression chamber and primary containment were declared inoperable on April 30, 2023, and a Technical Specification required shutdown was initiated.

Corrective Actions: PSEG initiated NOTF 20934082 to document inspector concerns, conducted rescreening and assigned corrective actions for NOTF 20919478.

Corrective Action References: 20934082, 20919478, 20935114

Performance Assessment:

Performance Deficiency: The inspectors determined that PSEGs failure to adequately identify and screen the NOTF as CAQ was a performance deficiency within their ability to foresee and correct and should have been prevented.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, PSEG did not properly identify and evaluate an adverse trend following the initial revelation of drywell to suppression chamber pressure trend during the normal nitrogen make up to the drywell, an indication of potentially bypass leakage, and corrective action was not assigned in accordance with PSEG corrective action program requirements. Consequently, the degraded condition went misdiagnosed from October 18, 2022 to April 30, 2023, and required a plant shutdown to perform repair and restore operability of the primary containment.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2 the finding screened to Green, very low safety significance, because the performance deficiency was not a design or qualification deficiency, did not involve in actual loss of safety function, or did not involve in loss of a probabilistic risk assessment function for greater than TS allowed outage time. The inspectors, in collaboration with the Region 1 senior reactor analysts, determined it was very low safety significance based on the following.

Additional details and bases are documented in this report in NCV 05000354/2023003-01, Improper Assembly and Inadequate Post-Maintenance Testing of the Vacuum Breaker Rendered the Primary Containment Inoperable.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, PSEG personnel deemed the initial NOTF operable, and then they relied on this assessment for subsequent occurrences, where they did not challenge the subsequent occurrences of drywell to suppression chamber pressure trend during their daily nitrogen make up, failing to identify the adverse trend that led to a degraded condition. This failure to properly identify and address the underlying issue demonstrates weakness in implementing an effective error reduction measure, such as avoiding complacency.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, specifies, 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.

PSEGs implementing procedure LS-AA-120, Issue Identification and Screening Process, Revision 24, directs all PSEG nuclear personnel and supplemental personnel to identify undesired conditions, including conditions adverse to quality, document them using a NOTF in the CAP, and ensure those conditions are corrected. Section 4.4.2 requires the SOC to classify each issue with appropriate categorization in accordance with the Attachment 2 of the procedure. Attachment 2 defines CAQ as any condition that describes a failure, malfunction, deficiency, defective item, or non-conformance in a safety-related structure, system, or component.

Procedure LS-AA-125, Corrective Action Program, Revision 29, Section 4.4.4 states, in part, Create a condition report corrective action (CRCA) in SAP for any planned action necessary to restore a CAQ.

Contrary to the above, from October 31, 2022 to April 19, 2023, PSEG did not promptly identify and assign corrective actions for an adverse trend that was a CAQ. This trend was apparent during their daily nitrogen make up to the drywell and challenged the primary containment integrity because it affected PSEG's capability to establish the initial differential pressure between the drywell and suppression chamber to demonstrate the operability of the primary containment.

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

Observation: Test Conditions for Drywell to Suppression Chamber Leak-by Test Could Not be Established 71152A The inspectors evaluated PSEGs corrective actions associated with PSEGs failure to establish the initial test conditions for the drywell to suppression chamber leak-by test. The issue and corrective actions completed were documented in PSEGs corrective action program as NOTFs 20935114, 20919478, 20934082, 20935223, 20932336, 20934975, 20934976, 20934983, and 20935324. PSEG completed RCE 70229351, which determined the direct cause of the issue was drywell to suppression chamber vacuum breaker A was not set appropriately to prevent mechanical interference from its associated limit switches and closing magnet assemblies. Related to this issue, PSEG determined the root cause was the refuel outage risk classification process that failed to identify first call PM on drywell to suppression chamber vacuum breaker as high risk. The outage management and risk management processes did not put in place measurable milestone requirements to ensure that the infrequently performed, LCO coded work scope was risk-screened by the Operations Outage Coordinator and received the required outage readiness challenge for high risk work.

The inspector reviewed PSEG's evaluations, WO documents, and planned and completed corrective actions, as applicable. The inspectors identified two Green findings associated with the A vacuum breaker degraded condition, as documented in this report.

EXIT MEETINGS AND DEBRIEFS

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

• On October 18, 2023, the inspectors presented the integrated inspection results to Robert DeNight, Site Vice President, and other members of the licensee staff.

• On August 10, 2023, the inspectors presented the IP 71124.05 radiation instrumentation exit debrief inspection results to Robert DeNight, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.07A Procedures

ER-AA-340-1002

Service Water Heat Exchanger and Component Inspection

Guide, Attachment 1, Heat Exchanger Data Sheet for H1-

EG-1A2E-201

10/11/2022

SH-PBD-AMP-

XI.M20, GALL

Program XI.M20

Open-Cycle Cooling Water Program

Work Orders

30342384