IR 05000354/2023001

Integrated Inspection Report 05000354/2023001

ML23116A003
Person / Time
Site:	Hope Creek
Issue date:	04/26/2023
From:	Brice Bickett NRC/RGN-I/DORS
To:	Carr E Public Service Enterprise Group
References
IR 2023001
Download: ML23116A003 (1)

Text

April 26, 2023

SUBJECT:

HOPE CREEK GENERATING STATION - INTEGRATED INSPECTION REPORT 05000354/2023001

Dear Eric Carr:

On March 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Hope Creek Generating Station. On April 19, 2023, the NRC inspectors discussed the results of this inspection with Tom Agster, Plant Manager, 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 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:

05000354/2023001

Licensee:

PSEG Nuclear, LLC

Facility:

Hope Creek Generating Station

Location:

Hancocks Bridge, NJ

Inspection Dates:

January 1, 2023 to March 31, 2023

Inspectors:

D. Beacon, Nuclear Engineer

J. Brand, Reactor Inspector

E. Eve, Senior Reactor Inspector

C. Hobbs, Reactor Inspector

J. Patel, Senior Resident 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

Failure to Correct an Emergency Diesel Generator (EDG) Jacket Water (JW) Flange Design Deficiency in a Timely Manner Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000354/2023001-01 Open/Closed

[H.10] - Bases for Decisions 71111.12 A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion XVI, Corrective Actions, was identified because PSEG did not correct C EDG JW flange design deficiency in a timely manner. Specifically, PSEG did not install longer bolts with locknuts or replace a gasket with a new type of material to resolve the known deficiency. As a result, the 'C' EDG experienced a leak from the JW supply flow orifice bolted flange connection during the 24-hour endurance test, which raised reasonable doubts of operability for the 'C' EDG to perform its specified safety function.

Human Error Results in Inoperable Residual Heat Removal (RHR) Shutdown Cooling Mode Isolation Actuation Instrumentation Channel Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000354/2023001-02 Open/Closed

[H.8] -

Procedure Adherence 71153 A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified because PSEG personnel failed to follow written procedure HC.IC-GP.ZZ-0115, Transmitter Isolation/Restoration Procedure, Sensitive Rack Instrumentation,

Instrument Rack 10C005 - RPV Channel C, Revision 15. Specifically, between October 5, 2022 and October 10, 2022, PSEG maintenance technicians failed to implement the specified procedure correctly when they installed an electrical jumper at a different location than described in the procedure. Due to this human error, the technicians inadvertently left the jumper in place, rendering one of the two-channel interlocks that protect the RHR system from over-pressurization inoperable.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000354/22-002-00 LER 22-002-00 for Hope Creek Generating Station,

Inoperable Isolation Actuation Instrumentation Caused by Failure to Remove Electrical Jumper 71153 Closed

PLANT STATUS

The Hope Creek Generating Station (Hope Creek) began the inspection period at rated thermal power. On February 25, 2023, Hope Creek was down powered to 70 percent to perform turbine valve testing, control rod pattern exchange, and remove the 'A' reactor feedwater pump from service for maintenance. The unit returned to rated thermal power on February 26, 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 (IP Section 03.01) (2 Samples)

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

(1) 'B' and 'D' core spray systems on January 24, 2023

(2) Reactor core isolation cooling system with the keep fill out of service for maintenance on March 14, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) Lower control equipment room in pre-fire plan FP-HC-3532 on January 30, 2023

(2) 'B' safety auxiliaries cooling system heat exchanger and pump room in pre-fire plan FP-HC-3432 on February 1, 2023

(3) Control building equipment mezzanine area in pre-fire plan FP-HC-3542 on March 17, 2023

(4) Class 1E switchgear rooms in pre-fire plan FP-HC-3541 on March 20, 2023

(5) Cable spreading room in FP-HC-3523 on March 20, 2023

Fire Brigade Drill Performance (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the on-site fire brigade training and performance during an unannounced fire drill on February 23, 2023

71111.06 - Flood Protection Measures

Flooding (IP Section 03.01) (1 Sample)

(1) Control building equipment mezzanine area on March 17, 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 valves testing and 'C' reactor feedwater pump maintenance on February 25, 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 a training exercise on February 7, 2023

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 remain capable of performing their intended function:

(1) Emergency diesel generators following a JW leak on February 13, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management (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) Planned inoperability of the 'C' EDG on January 11, 2023

(2) Planned inoperability of reactor core isolation cooling system on February 21, 2023

(3) Protected equipment walkdown for 'A' and 'C' core spray pump testing, 'E' filtration, recirculation, and ventilation system maintenance, and emergent work associated with reactor manual control system power supply issue on March 28, 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) 'C' EDG following identification of JW leakage during a surveillance test on January 20, 2023

(2) 'R' safety relief valve pilot temperature exceeding baseline monitoring temperature on March 6, 2023

(3) 'B' EDG output breaker lockout (86T) tripped during the output breaker remote operation functional test on March 13, 2023

(4) 'B' core spray pump delayed start relay found out of tolerance on March 15, 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 (IP Section 03.01) (4 Samples)

(1) 'A' EDG preventive maintenance window during the week of January 30, 2023

(2) 'A' core spray motor operated valve F015A following preventive maintenance on March 6, 2023

(3) Reactor manual control system following power supply issue due to a failed capacitor on March 28, 2023

(4) Scram discharge volume vent and drain valves solenoid replacement following indication of intermittent operation on March 29, 2023

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) HC.OP-ST.KJ-0016, 'C' EDG 24-hour operability run on January 11, 2023

(2) HC.OP-IS.BJ-0001, high pressure coolant injection main and booster pump quarterly test on March 9, 2023

(3) HC.OP-ST.KJ-0014, 'A' EDG 24-hour operability run and hot restart test on March 1, 2023

Inservice Testing (IP Section 03.01) (1 Sample)

(1) HC.OP-IS.BD-0001, reactor core isolation cooling pump - OP203 - Inservice test on February 22,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

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

(1) January 1, 2022 through December 31, 2022

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

(1) January 1, 2022 through December 31, 2022

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

(1) January 1, 2022 through December 31, 2022

===71152A - Annual Follow-up Problem Identification and Resolution

Annual Follow-up of Selected Issues (Section 03.03)===

(1) Root cause evaluations (70226537 and 70226259) for reactor coolant system chemistry excursion during the fall 2022 refueling outage.

(2) Review of corrective actions taken by PSEG as result of the control rod blade 42-59 bulging event that was observed during the fuel cycle 24 refueling outage in October 2022.

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) Licensee Event Report 05000354/2022-002-00, Inoperable Isolation Actuation Instrumentation Caused by Failure to Remove Electrical Jumper (ADAMS Accession No. ML23018A201). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is Closed.

INSPECTION RESULTS

Failure to Correct an Emergency Diesel Generator (EDG) Jacket Water (JW) Flange Design Deficiency in a Timely Manner Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems

Green NCV 05000354/2023001-01 Open/Closed

[H.10] - Bases for Decisions 71111.12 A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion XVI, Corrective Actions, was identified because PSEG did not correct C EDG JW flange design deficiency in a timely manner. Specifically, PSEG did not install longer bolts with locknuts or replace a gasket with a new type of material to resolve the known deficiency. As a result, the 'C' EDG experienced a leak from the JW supply flow orifice bolted flange connection during the 24-hour endurance test, which raised reasonable doubts of operability for the 'C' EDG to perform its specified safety function.

Description:

Hope Creek Generating Stations (Hope Creek) class 1E alternate current power system consists of four EDGs (A, B, C, and D), which serve as standby power supplies for the four safety-related 4.16 kV buses in the event of loss of normal power sources to the respective buses. The standby power supply for each 4.16 kV bus consists of one EDG complete with its auxiliaries, including the cooling water, starting air, lubrication, intake and exhaust, and fuel oil systems. The EDGs safety-related cooling water system consists of two separate cooling loops: the JW cooling loop and the intercooler cooling loop. The JW cooling loop circulates demineralized water with a corrosion inhibitor to control the operating temperature of the diesel engine by removing heat from the EDG cylinder jackets, turbocharger, and speed governor oil. Water inventory is maintained within the established limits of the system by an expansion tank connected to the pump suction piping. Protective devices and alarms to operators are installed on the EDG for JW temperature high, JW pressure, etc., to protect from catastrophic failure of the diesel and aid in operator response.

Under normal plant operation, makeup water to the expansion tank is supplied through an auto makeup valve from the non-safety-related demineralized water system. Under accident conditions, makeup to the JW through the non-safety-related demineralized water system is not considered. The Hope Creek updated final safety analysis report states that the EDG cooling water system is designed to seismic category 1 requirements and remains functional during and after a safe shutdown earthquake. Additionally, it is designed to withstand wind, tornadoes, floods, and missiles and to permit testing and inspection of active system components during plant operation.

On January 10, 2023, approximately 20 minutes into the performance of a C EDG 24-hour endurance surveillance test, in accordance with the procedure HC.OP-ST.KJ-0016, EDG 1CG400 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Operability Run and Hot Restart Test, Revision 37, an equipment operator in the field identified JW leaking at a rate of approximately 1900 mL/min from the C EDG JW supply header flow orifice bolted flange connection near cylinder #6. PSEG documented this condition in a notification (NOTF) 20924905. As a result of the leak, the surveillance test was stopped, and the diesel JW system was drained to perform a repair.

Following the repair, the surveillance test was re-started from the beginning on January 11, 2023. During this second surveillance test, a separate, smaller, leak developed on a flexible hose that was estimated to be approximately 30 to 75 mL/min (NOTF 20924911). PSEG mitigated this smaller leak to support the completion of the in-progress surveillance test, and declared the surveillance test completed satisfactorily. Subsequently, PSEG removed the C EDG from service, and repaired the flexible hose leak.

PSEG Operations performed an operability screening of the 'C' EDG degraded flange condition in NOTF 20924905, in accordance with OP-AA-108-115, Operability Determination, and documented that 'C' EDG was operable. The basis for operability, in part, was the operators ability to implement a compensatory action in accordance with HC.OP-ST.KJ-0001, to provide makeup water to the JW expansion tank in conditions where a leak reduces volume in the tank and challenges the EDG to perform its safety function.

The inspectors reviewed this event to assess the operability of the EDG and PSEG's response to address the degraded conditions. The inspectors noted that a similar JW flange leak occurred in September 2021 on the A EDG, which was the subject of a Green finding in Hope Creek inspection report 05000354/2021004. Therefore, the inspectors reviewed corrective actions that were performed as a result of the September 2021 leak, as they applied to the JW flange leak conditions that were observed on the C EDG on January 10, 2023.

The inspectors reviewed PSEG procedure LS-AA-125, "Corrective Action Program," Revision 28, to determine whether PSEG correctly implemented their corrective action process following the September 2021 JW leak on A EDG. PSEG performed an equipment reliability evaluation (ERE) 70219857 on November 3, 2021. This ERE concluded that the direct cause of the A EDG JW flange leak was loss of the minimum required applied gasket stress (force holding the gasket together) via bolt loosening and/or gasket deformation, and that this phenomenon was caused by inadequate design on the EDG flow orifice bolted flange connections which allows bolt loosening. The inspectors noted that this design deficiency was identified to exist in a total of eight flanges: two flanges on each of Hope Creeks EDGs.

As a result of the conclusions in the ERE, PSEG planned several corrective actions, typically referred to as a CRCA in PSEGs corrective action program (CAP) procedures and software, to address the condition adverse to quality that each applied to all four EDGs. Per LS-AA-125, CRCAs are to be completed within 180 calendar days. The first CRCA (70219857-0090)consisted of immediate torque checks on all eight of the identified susceptible flanges. This was completed successfully on January 3, 2022, and confirmed that the bolt loosening phenomena had been occurring on the C EDG. Specifically, the as found flange bolt torques on C EDG were determined to be ~15 ft-lb, when the specified torque requirement was 35 ft-lb. This corrective action returned the bolt torques to 35 ft-lb but did not address the design deficiency that caused the bolt loosening phenomenon.

The second CRCA (70219857-0080) that resulted from the November 2021 ERE was to create and implement preventive maintenance work orders to perform flange bolt re-torquing on a quarterly basis. These quarterly re-torque activities were completed throughout 2022 on the C EDG JW flanges.

The third corrective action (70219857-0070) that resulted from the November 2021 ERE was to replace the bolts in each of the susceptible flanges with longer bolts, and to install locknuts on the additional threaded length of the bolts, behind the flanges, to prevent the bolts from loosening. This corrective action was developed to correct the design deficiency identified in the ERE. However, PSEG planned this corrective action as a Long-Term Corrective Action (LTCA) with an initial planned completion target of August 1, 2022. The fourth corrective action (70219857-0110) assigned from the ERE was to replace the gasket in flow orifice flanges with a new improved material (identified as material master 1090994) on each of the EDGs. PSEG planned this action as an action tracking item and assigned equipment reliability CAP code, with an initial planned completion date of August 1, 2022. Per PSEG procedure LS-AA-125, Step 2.1.1, action tracking items are considered outside the CAP scope.

PSEG procedure LS-AA-125, Step 4.4.7, states, in part, DOCUMENT the justification and approval for designation as a LTCA in the SAP order and subsequently provides a note that specifies, in part, the justification for an LTCA created in an approved management review committee evaluation product is credited by the management review committee approved evaluation. Inspectors reviewed the November 2021 ERE, which created the LTCA and was approved by the PSEG management review committee. The inspectors determined that the ERE did not provide justification to support LTCA status for this third corrective action and did not support the extended planned completion target. Additionally, LS-AA-125, Step 4.4.5, states, in part, CRCAs that have initial target dates greater than 180 days may be considered for LTCA status if they meet the criteria outlined below and have obtained Plant Manager approval and subsequently includes a list of seven criteria regarding the feasibility of planning and performing the corrective action. The inspectors reviewed each of these criteria as they applied to the planned corrective action of installing longer bolts and locknuts on the EDG JW flanges. Inspectors concluded that the corrective action did not reasonably meet the criteria specified, and therefore the LTCA status should not have been applied when it was created under the ERE.

The inspectors noted that the third corrective action from the ERE was designated LTCA status and originally planned to be completed by August 1, 2022, which is outside the typical timeline (~180 days) for a CRCA status corrective action. Additionally, management review committee approved a due-date extension of 1 year on July 28, 2022, pushing the corrective action's target completion date to August 1, 2023. The inspectors noted that the fourth corrective action assigned as an action tracking item with CAP code was also extended, consistent with a date change of LTCA. Additionally, the inspectors noted that a common-cause evaluation (70134049) from 2012 performed to address the A EDG JW leak from a bolted flange connection had a corrective action assigned as an LTCA to upgrade the EDG cork and vegetable gasket to Garlock material. This item was not completed and closed in 2015.

As a result, the inspectors concluded that the corrective actions were inappropriately planned in an untimely fashion and therefore not performed prior to additional JW leaks. The C EDG was therefore subject to the identified design deficiency for an extended period until it eventually developed a JW flange leak during the surveillance test on January 10, 2023.

Corrective Actions: PSEGs immediate corrective action was to replace the degraded JW flange gasket and re-perform the 24-hour EDG surveillance test. PSEG then declared the EDG operable following repair of the smaller, unrelated, flexible hose leak on January 12, 2023. Subsequently, on January 19, 2023, as part of follow-up inspection activities, the inspectors identified that the C EDG JW flange repair was completed using a new gasket material (a planned upgrade) and the original bolts. Specifically, the third corrective action to install longer bolts and locknuts had not been implemented as part of the repair, even though the JW flange bolts were completely removed, and the longer bolts and locknuts were available on-site at the time. Additionally, the inspectors identified that the original bolts were not fully engaged in the JW flange and did not have the minimum required protruding threads required by PSEG maintenance practices. PSEG captured the condition in NOTF 20925707 and ultimately removed the EDG from service on January 20, 2023, to implement the corrective action using longer bolts and locknuts. The EDG was then returned to service following a post-maintenance test run. Subsequently, PSEG completed a technical evaluation 70227603-0060 and concluded that the available thread engagement in the C EDG JW piping flanges noted in NOTF 20925707 is considered to have been structurally adequate for the imposed design loads during their time of installation. The shortened thread engagement did not reduce the ability of the JW piping from performing its design function.

Corrective Action References: 20924905, 20925707, 20924911

Performance Assessment:

Performance Deficiency: The inspectors determined that inadequate planning and implementation of corrective actions to address an identified design deficiency related to Hope Creek's EDG JW flanges was a performance deficiency that was reasonably within PSEGs ability to foresee and correct.

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, the degraded JW flange connection resulted in leakage exceeding PSEGs established acceptable leak rate for purposes of initially assessing operability, and implementation of a manual compensatory action would have been required in a short period of time to ensure the diesel would be able to perform its specified safety function. Additionally, the inspectors referenced the NRC IMC 0612, Appendix E, Examples of Minor Issues, and determined that examples 3.k, 3.m and 3.g were similar, and informed more than minor determination because the condition resulted in reasonable doubt of the operability of the C EDG, and additional actions were necessary to support operability.

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, "Mitigating System Screening Questions," the inspectors determined that the deficiency affected the design of the C EDG JW bolted flange connection. However, with the PSEGs ability to implement a proceduralized manual action in a timely manner, which provides makeup water to the EDG JW expansion tank from a safety-related water source, would have allowed the C EDG to maintain its probabilistic risk assessment functionality.

Thus, the finding was screened to Green.

Cross-Cutting Aspect: H.10 - Bases for Decisions: Leaders ensure that the bases for operational and organizational decisions are communicated in a timely manner. Specifically, PSEG management failed to recognize that corrective actions to address a design deficiency affecting multiple EDG trains were planned on an inappropriate timeline that was not adequately justified, and subsequently approved and further delayed those untimely corrective actions.

Enforcement:

Violation: Title 10 of the Code of Federal Regulations, Part 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.

Contrary to the above, from November 3, 2021 to January 20, 2023, PSEG failed to promptly correct a design deficiency affecting the C EDG JW system. As a result, during surveillance testing on January 10, 2023, the C EDG developed a significant JW leak that called into question the diesels capability and reliability to respond to a design-basis event.

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

Observation: Control Rod Blade 42-59 Bulging 71152A In May 2021, during the reactor startup for fuel cycle 24, the control rod in peripheral core location 42-59, did not respond to attempts to withdraw the rod from the control room.

Abnormal operating procedure actions were attempted to move the control rod from the fully inserted position which included raising the control rod drive mechanism drive water differential pressure. Additional operator actions were not successful, and control rod 42-59 remained fully inserted for the rest of the reactor startup, and all of fuel cycle 24. With the control rod in the fully inserted position, the control rod performed its intended safety function to add negative reactivity to the core. In addition, this control rod underwent normal rod testing activities prior to reactor startup in May 2021, and functioned normally during control rod exercising, normal speed testing, and scram time testing.

In October 2022, during refueling outage 24 (H1R24), two fuel bundles adjacent to control rod 42-59 at core locations 43-58 and 43-60, were unable to be removed from the core during refueling operations. Camera inspection revealed bulging in one wing of control rod 42-59, at the bottom of the blade, near the core plate. The bulging resulted in the blade coming into contact with the fuel channeling on these two fuel bundles, which created a frictional force that overloaded the refueling bridge hoist, resulting in the hoist automatically stopping the lifting operation. Different troubleshooting efforts were attempted. Ultimately, the refuel bridge hoist PLC logic was reprogrammed to allow a faster raising speed that removed fuel bundle 43-58 from its fuel support piece on the core plate. Once fuel bundle 43-58 was discharged to the spent fuel pool, fuel bundle 43-60 and control rod 42-59 were also removed from the core and discharged to the spent fuel pool. In November 2022, power ascension operations were completed, and Hope Creek entered fuel cycle 25.

Apparent Cause Evaluation (ACE) 70225969 was conducted in December 2022, to examine and understand the causes of the control rod blade 42-59 bulging event. The direct cause of the bulging on control blade 42-59 was determined to be water intrusion into the top of the blade from cracking. The cracking was caused by irradiated assisted stress corrosion cracking, which is known to occur in control rod blades that reach a high enough neutron exposure in the core. Through gravity, water traversed the length of the control rod blade, into the bottom of the blade. During reactor startup, the water flashed to steam causing a bulge on both faces of one of the four wings of the control rod blade.

As part of the ACE corrective actions, PSEG reviewed the most recent control rod blade vendor guidance promulgated in 2016, and determined that this vendor guidance had been captured in PSEG procedure, NF-AB-135-1410, "BWR Control Blade Lifetime Management,"

Revision 3, Attachment 7, to this procedure titled, "Inspection Criteria for CR-82M, CR-82M-1, and CR-82M-2, contains guidance on the lifetime management recommendations for Westinghouse model CR-82M control rod blades. The basic strategy is to perform visual inspection of control rod blades that reach a certain percent B-10 depletion value, to determine if irradiated assisted stress corrosion cracking developed in the stainless steel sheaths on the wings. If cracks are observed, the control rod blade is permanently discharged to the spent fuel pool. If a control rod blade exceeds the recommended percent B-10 depletion value, it cannot be re-used in a power shaping location. This guidance was followed for rod 42-59, however, the control rod failed prematurely, after passing a satisfactory visual inspection in fuel cycle 16, and before exceeding percent B-10 burnup criteria. As a result, PSEG determined the apparent cause of the control rod blade 42-59 bulging event in H1R24 was inadequate vendor guidance for Westinghouse CR-82M control rod blades.

Immediate corrective actions taken by PSEG were to permanently discharge control rod 42-59 to the spent fuel pool, along with the two adjacent fuel bundles, 43-58 and 43-60 in H1R24. Additional planned corrective actions include revising NF-AB-135-1410, "BWR Control Blade Lifetime Management," Attachment 7, to ensure the guidance in this section is adequate, working with the control rod blade vendor to incorporate the OE from H1R24 into the vendor guidance for CR-82M control rod blades, and evaluate the OE from H1R24 for Part 21 reportability. In Hope Creek fuel cycle 25, there are 62 Westinghouse CR-82M model control rod blades installed in the core. Six blades are in power shaping locations, while the other 56 control rods are in shutdown core locations, and are expected to be fully withdrawn during fuel cycle 25. PSEG plans to remove all Westinghouse CR-82M model control rod blades over the next two refueling outages.

The inspectors interviewed PSEG nuclear fuels engineers, reactor engineering staff, and reviewed the planned corrective actions detailed in ACE 70225969. The inspectors did not identify any findings or violations of more than minor significance.

Observation: Root Cause Evaluations (70226537 and 70226259) for Reactor Coolant System Chemistry Excursion During the Fall 2022 Refueling Outage 71152A The licensee performed two root cause evaluations to assess the introduction of high conductivity water from the condensate system to the reactor coolant system on October 17, 2022, during startup activities. The inspectors focused on the root causes, corrective actions, and extent of condition of the transport of high conductivity water to the reactor vessel. The licensee determined the root cause to be that chemistry did not identify or implement an effective method to ensure samples were taken to support condensate system startup in accordance with CY-AB-120-110, Condensate and Feedwater Chemistry. To correct this issue, the licensee created an action item to update the chemistry outage sampling plan to include direction for placing the balance of plant sample sink in service and taking condensate samples as directed by CY-AB-120-110. The licensee also created an action item to revise HC.OP-SO.AD-0001, Condensate System Operation, to include steps directing chemistry to perform primary condensate pump discharge samples in accordance with CY-AB-120-110 after placing the first primary condensate pump in service. Similarly, the licensee created an action item to revise HC.OP-SO.DA-0001, Circulating Water System Operation, to perform primary condensate pump discharge samples in accordance with CY-AB-120-110 after placing the first circulating water pump in service.

The inspectors reviewed the root cause evaluations and noted that the licensee did not implement the sampling requirements in CY-AB-120-110, Condensate and Feedwater Chemistry, during plant startup activities on October 17, 2022. The inspectors determined the failure to implement the procedure was minor because it did not adversely affect the barrier integrity cornerstone objective to ensure that physical barriers protect the public from radionuclide releases caused by accidents or events. The inspectors noted that the conductivity and chlorides values in updated final safety analysis report, Table 5.2-8, for the reactor coolant system were exceeded during the event. The inspectors reviewed the engineering evaluation and determined the structural integrity of the reactor vessel was not affected because the chemistry excursion event occurred under cold shutdown conditions where environmentally assisted cracking was not a concern. PSEG captured this observation in notification 20930847. This observation does not constitute a more than minor violation in accordance with the Enforcement Policy.

Human Error Results in Inoperable Residual Heat Removal (RHR) Shutdown Cooling Mode Isolation Actuation Instrumentation Channel Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events

Green NCV 05000354/2023001-02 Open/Closed

[H.8] -

Procedure Adherence 71153 A self-revealing Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified because PSEG personnel failed to follow written procedure HC.IC-GP.ZZ-0115, Transmitter Isolation/Restoration Procedure, Sensitive Rack Instrumentation, Instrument Rack 10C005 - RPV Channel C, Revision 15. Specifically, between October 5, 2022 and October 10, 2022, PSEG maintenance technicians failed to implement the specified procedure correctly when they installed an electrical jumper at a different location than described in the procedure. Due to this human error, the technicians inadvertently left the jumper in place, rendering one of the two-channel interlocks that protect the RHR system from over-pressurization inoperable.

Description:

The Hope Creek technical specification Limiting Condition for Operation (LCO)3.3.2, requires the isolation actuation instrumentation channels in Table 3.3.2-1, trip function 7.b for RHR system shutdown cooling isolation operable while the unit is in operational condition (OPCON) 1, 2, or 3. The reactor vessel (RHR cut-in permissive) pressure high trip function 7.b of Table 3.3.2-1, protects the RHR system from over-pressurization by providing two-channel interlocks that prevent opening RHR shutdown cooling valves when reactor pressure is above 82 psig. RHR system valves interlocked by this function are inboard and outboard RHR shutdown cooling suction valves (HV-F009 and HV-F008), and shutdown cooling return valves (HV-F015A and HV-F015B).

On October 5, 2022, while the facility was in OPCON 5, refueling, PSEG maintenance technicians performed a preventive maintenance (work order 30358492) on reactor pressure vessel (RPV) level and pressure transmitters located on the RPV channel C instrument rack 10C005. Procedure HC.IC-GP.ZZ-0115, Revision 15, was used for isolation and returning to service the transmitters on instrument rack 10C005 with common sensing lines. Specifically, this procedure provided instructions to disable system trips and initiations associated with these transmitters, and provided directions to install electrical jumpers at specific locations to preclude an inadvertent engineered safety feature actuation while performing valve manipulation of those transmitters. Step 5.3.12 of this procedure, required installing a jumper between terminals KKK-9 and KKK-10, in panel H11-P609, to prevent RHR shutdown cooling valve isolation due to thermal expansion, while the RPV pressure transmitter (B21-N078C)was isolated. The technicians performed this step by placing a jumper across test jacks KKK-9 and KKK-10, which was different from terminal points KKK-9 and KKK-10. The inspectors determined that the failure to install a jumper according to the procedure led to a human error on October 10, 2022, when technicians performed Step 5.5.15, which required the removal of jumpers from the terminal point KKK-9 and KKK-10. Technicians marked Step 5.5.15 as not applicable because they did not find any jumpers installed between terminal point KKK-9 and KKK-10. As a result of this human error, the jumper remained installed in the plant when the facility transitioned to OPCON 2, startup, on October 28, 2022, and OPCON 1, power operation, on October 30, 2022.

On November 30, 2022, during the performance of the technical specification surveillance test in accordance with procedure HC.IC-CC.SM-0012, NSSSS - Division 4 Channel B21-N681D Reactor Vessel Level (Trip 1, 2), Revision 23, at Step 4.1.26.2, PSEG maintenance technician discovered that a jumper was already installed at the location of test jacks KKK-9 and KKK-10 in panel H11-P609. They did not expect to find any electrical jumper in this panel and had not been briefed on this configuration. As a result, PSEG completed a prompt investigation (NOTF 20922956) and a workgroup evaluation (70226871), which is a type of CAP causal evaluation, to determine the cause of this issue. PSEG's investigation revealed that the direct cause of the incident was human error when the technician incorrectly marked a required step as not applicable while performing HC.IC-GP.ZZ-0115, Step 5.5.15, on October 10, 2022.

PSEG determined the jumper remained installed from October 28, 2022 to November 30, 2022. During this period, the Hope Creek technical specification LCO 3.3.2 was applicable, and Table 3.3.2-1, trip function 7.b, required two-channel interlock for RHR shutdown cooling isolation valves operable. The jumper rendered one channel of two-channel interlock inoperable. As a result of this issue, PSEG submitted LER 05000354/2022-002-00 in accordance with Title 10 of the Code of Federal Regulation 50.73(a)(2)(i)(B), as required for the condition prohibited by technical specification. This non-cited violation serves as closure documentation for LER 05000354/2022-002-00.

Corrective Actions: PSEGs immediate corrective actions were to remove the jumper and restore two-channel interlock that protects RHR system from over-pressurization to operable on November 30, 2022, and perform prompt investigation. Additionally, procedure HC.IC-GP.ZZ-0115 was revised to provide direction to use test jack KKK-9 and KKK-10 when installing and removing jumpers in Steps 5.3.12 and 5.5.15, respectively.

Corrective Action References: 20922956, 20931449

Performance Assessment:

Performance Deficiency: The inspectors determined that PSEGs failure to follow procedure HC.IC-GP.ZZ-0115, Transmitter Isolation/Restoration Procedure, Sensitive Rack Instrumentation, Instrument Rack 10C005 - RPV Channel C, Revision 15, as written, was a performance deficiency because it was within PSEGs 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 Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the electrical jumper remained installed during OPCON 1, 2, and 3 at reactor pressure greater than 82 psig, which reduced the redundancy of a two-channel interlock to a single channel that protects the RHR system from over-pressurization and provides protection to prevent the intersystem loss of coolant accidents (LOCA), and the cornerstone objective was therefore affected. Additionally, the inspectors referenced the NRC IMC 0612, Appendix E, Examples of Minor Issues, and determined that example 6.c was similar, and informed this more than minor determination.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 1, Initiating Events Screening Questions, the finding screened to Green, very low safety significance, because after a reasonable assessment of degradation, the finding would not have likely resulted in exceeding the reactor coolant system leak rate for a small break LOCA, and would not have likely affected systems used to mitigate a LOCA or would not have likely caused intersystem LOCA. The redundant isolation actuation instrument channel was operable throughout this event to provide over-pressure protection for the RHR shutdown cooling piping, and the affected valves remained closed for the entire event.

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, the maintenance technicians did not follow written procedure steps correctly when installing an electrical jumper. As a result, a jumper remained installed in the plant, which affected the technical specification operability of a safety-system isolation instrumentation channel.

Enforcement:

Violation: Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with these instructions, procedures, or drawings. PSEGs procedure HC.IC-GP.ZZ-0115, Transmitter Isolation/Restoration Procedure, Sensitive Rack Instrumentation, Instrument Rack 10C005 - RPV Channel C, Revision 15, implemented this requirement to provide instructions for isolation and returning to service the safety-related transmitters on instrument rack 10C005 with common sensing lines. Additionally, the Hope Creek technical specification LCO 3.3.2 requires that the isolation actuation instrumentation channel shown in Table 3.3.2-1 shall be operable. Table 3.3.2-1, trip function 7.b, Reactor Vessel (RHR Cut-in Permissive) Pressure - High, requires two operable channels per trip system in OPCON 1, 2, and 3, and when this LCO is not met, action B requires placing the inoperable channel in the tripped condition within twelve hours or taking the action required by Table 3.3.2-1. For trip function 7.b, Table 3.3.2-1, refers to action 27. Action 27 requires the affected system isolation valves to be locked closed within one hour, and the affected system is declared inoperable.

Contrary to the above, PSEG maintenance personnel did not perform safety-related activities between October 5, 2022 and October 10, 2022, in accordance with the procedure HC.IC-GP.ZZ-0115, Revision 15, because they did not implement the procedure as written.

Specifically, PSEG maintenance technicians installed an electrical jumper at a different location than what was described in the procedure step. As a result of this human error, the technicians inadvertently left the jumper in place, rendering one of the two-channel interlocks that protect the RHR system from over-pressurization inoperable from October 28, 2022 to November 30, 2022, a period of 33 days.

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

EXIT MEETINGS AND DEBRIEFS

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

• On February 10, 2023, the inspectors presented the control rod blade 42-59 bulging event inspection results to Robert DeNight, Site Vice President, and other members of the licensee staff.

• On March 16, 2023, the inspectors presented the annual problem, identification and resolution sample results for inspection of root cause evaluations associated with the reactor coolant system chemistry excursion in October 2022 inspection results to Tom Agster, Plant Manager, and other members of the licensee staff.

• On April 19, 2023, the inspectors presented the integrated inspection results to Tom Agster, Plant Manager, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.24

Corrective Action

Documents

Resulting from

Inspection

20924884*

71152A

Corrective Action

Documents

225969

Apparent Cause Evaluation for control blade 42-59 bulge

2/12/2022

NOTFs

20877108,

20917723,

20917759,

20919828,

20918956

Corrective Action

Documents

Resulting from

Inspection

NOTF 20927639

Engineering

Changes

SAP 80129619

Core Operating Limits Report for Hope Creek Generating

Station Reload 24 / Cycle 25

10/18/2022

Engineering

Evaluations

NSF 21-038

Hope Creek Cycle 24 Core Loading Plan

04/27/2021

NSF 22-031

Control Blade Replacement Scope for Hope Creek RF24

through RF27

08/25/2022

NSF 22-069

Hope Creek Cycle 25 Core Loading Plan

09/22/2022

Procedures

HC.RE-FR.ZZ-

0002(Q)

Control Rod Removal and Installation

Revision 34

NF-AA-803

Nuclear Fuels Vendor Information Processing

Revision 0

NF-AB-135-1410

BWR Control Blade Lifetime Management

Revision 3

NF-HC-310-1002

Shuffle Works User's Guide

Revision 3