Text

Integrated Inspection Report 05000354/2021003
	ML21305A011
Person / Time
Site:	Hope Creek
Issue date:	11/02/2021
From:	Brice Bickett; NRC/RGN-I/DORS
To:	Carr E; Public Service Enterprise Group
References
	IR 2021003
	Download: ML21305A011 (15)
	v • d • e

November 2, 2021

SUBJECT:

HOPE CREEK GENERATING STATION - INTEGRATED INSPECTION REPORT 05000354/2021003

Dear Mr. Carr:

On September 30, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Hope Creek Generating Station. On October 19, 2021, the NRC inspectors discussed the results of this inspection with Mr. Steve Poorman, Hope Creek Plant Manager, and other members of your staff. The results of this inspection are documented in the enclos ed report.

One finding of very low safety significance (Green) is documented in this report.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555- 0001; with copies to the Regional Administrator, Region 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 Reactor Projects Branch 3 Division of Operating Reactor Safety

Docket No. 05000354 License No. NPF-57 E. Carr 2

Enclosure:

As stated

Inspection Report

Docket Number: 05000354

License Number: NPF-57

Report Number: 05000354/2021003

Enterprise Identifier: I-2021-003- 0000

Licensee: PSEG Nuclear, LLC

Facility: Hope Creek Generating Station

Location: Hancocks Bridge, NJ

Inspection Dates: July 01, 2021 to September 30, 2021

Inspectors: D. Beacon, Resident Inspector P. Finney, Senior Project Engineer M. Hardgrove, Resident Inspector B. Lin, Nuclear Systems Engineer J. Patel, Senior Resident Inspector

Approved By: Brice A. Bickett, Chief Reactor Project Branch 3 Division of Operating Reactor Safety

Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Hope Creek Generating Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Inadequate Risk Assessment of a Failed Low-Pressure Suction Switch Configuration Resulting in a Trip of the B Reactor Feedwater Pump Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.5] - Work 71111.12 FIN 05000354/2021003- 01 Management Open/Closed A self-revealed Green finding of OP-AA-107, Integrated Risk Assessment, Revision 1, was identified because PSEG did not perform a risk assessment of the condition following a failure of the low-pressure suction switch on the B reactor feedwater pump (RFP) on June 14, 2021.

Specifically, PSEG did not perform an adequate risk assessment that supported remaining in a configuration with the failed pressure switch energized after discovering water inside that switch. Subsequently, on June 16, 2021, a short to ground fault occurred inside that switch, which was cleared by opening a fuse in the trip circuit, causing a loss of power to the 'B' RFP trip circuit and a trip of the 'B' RFP. Hope Creek plant automatically responded to the loss of feed pump by initiating 'A' and 'B' reactor recirculation pumps (RRP) intermediate runback and power reduction to approximately 71 percent reactor thermal power (RTP).

Additional Tracking Items

None.

PLANT STATUS

The Hope Creek Generating Station (Hope Creek) began the inspection period at approximately 30 percent rated thermal power (RTP) coming out of a maintenance outage that replaced A safety relief valve (SRV) assembly following indications of pilot valve leakage. On July 21, Hope Creek commenced a shutdown for a maintenance outage to replace the H SRV following indications of pilot valve leakage. Following the corrective maintenance, Hope Creek returned to 98 percent RTP on July 25, and reduced power to approximately 72 percent RTP on July 26 to address elevated temperature on B reactor recirculation pump seal #2. Hope Creek returned to approximately 100 percent RTP on July 29. On September 17, Hope Creek commenced a planned load reduction to approximately 62 percent RTP to perform turbine valve testing, control rod pattern exchange, and corrective maintenance on A reactor feedwater pump. Power was restored to 100 percent RTP on September 19. The station remained at or near 100 percent RTP 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 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.

Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), resident and regional inspectors were directed to begin telework and to remotely access licensee information using available technology. During this time, the resident inspectors performed periodic site visits each week, increasing the amount of time on-site as local COVID-19 conditions permitted.

As part of their on-site activities, resident inspectors conducted plant status activities as described in IMC 2515, Appendix D, Plant Status, and conducted routine reviews using IP

===71152, Problem Identification and Resolution, observed risk significant activities; and completed on-site portions of IPs. In addition, resident and regional baseline inspections were evaluated to determine if all or a portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections were performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on-site.

The inspections documented below met the objectives and requirements for completion of the IP.

REACTOR SAFETY

71111.01 - Adverse Weather Protection

Seasonal Extreme Weather Sample (IP Section 03.01)===

Main article: IP 71152

(1) The inspectors evaluated the stations summer readiness on July 8

External Flooding Sample (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the stations readiness to cope with external flooding on August 17 through 20

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (4 Samples)

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

(1) Fire protection system on July 14

(2) High pressure coolant injection system on August 24

(3) A and C station service water system on August 31

(4) 'B' and 'D' 4.16 kV safety related buses on September 21

71111.05 - Fire Protection

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

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

(1) Fire water pump house in pre-fire plan FP-HC-3714 on July 14

(2) Control equipment mezzanine in pre-fire plan FP-HC-3542 on August 2

(3) Filtration, recirculation, and ventilation system, spent fuel pool and gamma scan detector area, recombiner area in pre-fire plan FP-HC-3461 on August 2

(4) Emergency diesel generator rooms and corridor in pre-fire plan FP-HC-3531 on August 25

(5) Electrical equipment area in pre-fire plan FP-HC-3513 on September 2

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 plant shutdown for a maintenance outage on July 21, and during reactor startup and plant startup on July 23

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)

(1) The inspectors observed and evaluated a crew of licensed operators in the plants simulator during a licensed operator requalification training on September 7

(2) The inspectors observed and evaluated a crew of licensed operators in the plants simulator during a licensed operator requalification training on September 13

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) Reactor feedwater pumps following multiple trips and equipment issues on July 9

(2) 'D' emergency diesel generator speed switch failure in June 2021 on July 29

(3) Control area ventilation and chilled water system following multiple system trips on August 11

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) Emergent work control for A filtration, recirculation and ventilation system vent fan declared inoperable following charcoal sampling on July 20

(2) Planned inoperability of reactor core isolation cooling system during battery charger maintenance window on August 24 through 26

(3) Emergent troubleshooting and repair of A moisture separator drain level controller on September 9

(4) 'A' emergency diesel generator during 24-hour surveillance test and following repair of jacket water flange leakage on September 27 through 29

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)

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

(1) H safety relief valve with indications of pilot stage leakage on July 12

(2) D emergency diesel generator crankcase oil moisture content greater than fault level discovered on March 10, 2021 on July 19

(3) B reactor feedwater pump functionality with low suction pressure trip disabled on August 3

(4) 'C' and 'D' emergency diesel generators rocker arm lube oil system high water content on August 4

(5) R safety relief valve following failure to lift when removed during the refueling outage on August 13

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) Temporary installation of supplementary cooling on B reactor recirculation pump seal purge flow lines on September 27

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (6 Samples)

The inspectors evaluated the following post-maintenance test activities to verify system operability and functionality:

(1) 'A' safety relief valve pilot replacement following indications of pilot leakage on July 7

(2) C emergency diesel generator 24-month preventive maintenance window on July 13

(3) H safety relief valve pilot replacement following indications of pilot leakage on July 22

(4) Reactor core isolation cooling system following direct current bus preventive maintenance on August 26

(5) Reactor water cleanup following automatic isolation due to failed fuse in the bailey cabinet on September 13

(6) A emergency diesel generator jacket water leak repair and retest on September 30

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) Maintenance outage for replacement of H safety relief valve replacement following indications of pilot leakage on July 22

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Surveillance Tests (other) (IP Section 03.01)

Main article: IP 71111.22

(1) HC.IC-FT.BB-0066 and HC.IC-FT.BB-0067, 'A' and 'B' reactor recirculation pumps seal cavity excess flow check valves on August 5

(2) HC.OP-IS.BJ-0001, high pressure coolant injection system quarterly surveillance test on September 7

(3) HC.OP-IS.BC-0003, 'B' residual heat removal pump in-service test on September 15

(4) HC.OP-ST.KJ-0014, 'A' emergency diesel generator 24-hour operability run and hot restart test on September 30

Inservice Testing (IP Section 03.01) (1 Sample)

(1) HC.OP-IS.BJ-0101, high pressure coolant injection system valves on July 13

FLEX Testing (IP Section 03.02) (1 Sample)

(1) SH.OP-PT.FLX-0480, FLEX 480V diesel generators on August 24

71114.06 - Drill Evaluation

Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)

(1) The inspectors observed an emergency preparedness drill on September

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)===

Main article: IP 71151

(1) July 1, 2020 - June 30, 2021 on August 4

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

(1) July 1, 2020 - June 30, 2021 on August 4

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

(1) July 1, 2020 - June 30, 2021 on August 5

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

(1) July 1, 2020 - June 30, 2021 on August 6

71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues (IP Section 02.03) (1 Sample)

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

(1) Review of events and corrective actions involving leak detection monitoring system low voltage power supply failures.

INSPECTION RESULTS

Inadequate Risk Assessment of a Failed Low-Pressure Suction Switch Configuration Resulting in a Trip of the B Reactor Feedwater Pump Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.5] - Work 71111.12 FIN 05000354/2021003- 01 Management Open/Closed A self-revealed Green finding of OP-AA-107, Integrated Risk Assessment, Revision 1, was identified because PSEG did not perform a risk assessment of the condition following a failure of the low - pressure suction switch on the B reactor feedwater pump (R FP) on June 14, 2021.

Specifically, PSEG did not perform an adequate risk assessment that supported remaining in a configuration with the failed pressure switch energized after discovering water inside that switch.

Subsequently, on June 16, 2021, a short to ground fault occurred inside that switch, which was cleared by opening a fuse in the trip circuit, causing a loss of power to the 'B' RFP trip circuit and a trip of the 'B' RFP. Hope Creek plant automatically responded to the loss of feed pump by initi ating 'A' and 'B' reactor recirculation pumps (RRP) intermediate runback and power reduction to approximately 71 percent reactor thermal power (RTP).

Description:

The reactor feedwater system at Hope Creek consists of three steam -driven RFPs, three high pressure feedwater heaters, and the necessary valves, piping, instrumentations, and controls required to deliver sufficient feedwater flow to the reactor pressure vessel during all modes of operations. There are three low - pressure suction switches per RFP that are configured in two out of three logic for tripping the RFP turbine on low-low suction pressure at 230 PSIG.

On June 14, 2021, the main control room rec eived an overhead alarm for the 'B' RFP suction pressure sensor failure. PSEG entered this condition in their corrective action program under notification (NOTF) 20879321 and implemented a troubleshooting plan in accordance with the procedure MA - AA-716 - 004, Conduct of Troubleshooting, Revision 15, to determine the cause of the alarm. PSEG identified through troubleshooting activities that a pressure switch (H1AE - 1AEPSLL - 1777B3) for the 'B' RFP low - pressure suction trip had failed. PSEG also discovered the pressure switch was full of water, which indicated a failure of diaphragm, and determined it required replacement. Implementing the troubleshooting plan, PSEG lifted the leads on the output of the pressure switch into the trip logic to reduce the risk o f an inadvertent 'B' RFP trip due to another low - pressure suction switch failure.

There were no spare switches available for replacement at the time.

Therefore, PSEG decided to stay in a troubleshooting plan and remained in the configuration with lifted leads on the output of the switch, but kept the failed low - pressure switch energized from the line side.

Subsequently, on June 16, 2021, due to the presence of water inside the switch, a short to ground fault blew the fuse in the 'B' RFP trip circuit.

This fuse provided power to all three low - pressure switches, as well as high reactor water level switches, and low flow switches.

When the fuse blew and deenergized the circuit, this caused the trip circuit to activate and trip the 'B' RFP, resulting in aut omatic plant response that initiated 'A' and 'B' RRPs intermediate runback and power reduction to approximately 71 percent RTP.

PSEG performed a prompt investigation (NOTF 20878950) and completed an equipment reliability evaluation (70218387) to determine the cause of the 'B' RFP trip.

The inspectors reviewed PSEG's evaluation and prompt investigation report, which identified the cause of the 'B' RFP trip was the blown fuse in Bailey 125 VDC cabinet. The cause of the blown fuse was short to ground fault from water intrusion in the low-pressure switch. The water was able to reach the switch due to a failed diaphragm.

The inspectors reviewed PSEG's procedural guidance for troubleshooting in the procedure MA-AA-716-004, Revision 15. Step 4.2.2 states that operational risk will be addressed using procedure WC-AA-105. The inspectors noted that procedure WC-AA-105 was not active at the time and was superseded by OP-AA-107, Integrated Risk Management, Revision 1.

OP-AA-107, Step 4.3.4, states, in part, that for all medium risk activities, the work activity owner or designee completes a Risk Management Plan (RMP) in accordance with 3 of the procedure.

The inspectors identified that PSEG did not complete a RMP in accordance with the Attachment 3 of OP-AA-107 prior to implementing a troubleshooting plan, and when continued with new configuration with the failed switch energized following troubleshooting. Therefore, the inspectors determined that PSEG was not in compliance with OP-AA-107, Step 4.3.4, because PSEG did not assess the risk associated with operating the pressure switch with failed diaphragm in an energized state and did not address its impact on the RFP trip circuit due to potential for a short to ground fault.

Corrective Actions: PSEGs immediate corrective action was to deactivate the RFP low-low suction pressure trip inputs for the B RFP turbine. Additionally, PSEG is evaluating a design change to eliminate the RFP low-pressure suction switches on all three pumps.

Corrective Action References: NOTFs 20878950, 20879321 and 20885221

Performance Assessment:

Performance Deficiency: The inspectors determined that failure to perform a risk assessment of a failed low-pressure suction switch configuration, which resulted in a trip of the 'B' RFP, was a performance deficiency that was reasonably within PSEG's ability to foresee 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, PSEGs failure to perform a risk assessment of the failed low-pressure suction switch and not addressing its impact on the RFP trip circuit due to a potential short to ground fault, resulted in blown fuse and trip of the B RFP, which caused an automatic plant response to initiate RRPs runback to reduce power to 71 percent RTP.

Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power.

Exhibit 1 - Initiating Event Screening Questions, Section B for Transient Initiators. The finding screened to Green, very low safety significance, because the B RFP trip did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.

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, PSEG did not implement a work process including the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, PSEGs work activity did not use the correct procedure to document the RMP through the work management process. [H.5]

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Observation : Missed Opportunities to Correct the Preventive Maintenance (PM) 71152 Strategy Deficiency The inspectors evaluated corrective actions taken in response to multiple Nuclear Measurement Analysis and Control (NUMAC) failures that occurred in recent years.

NUMAC failures increased in frequency in recent years and caused multiple isolations of the reactor water cleanup system, which prompted PSEG to take further actions. Specifically, two equipment reliability evaluations (EREs) were performed beginning in 2018.

The EREs examined historical performance since 2007 as well as recent specific failures. Many of the failures were originally thought to be related to specific circuit cards, in many cases the A10 card. Ultimately, in 2020, PSEG determined, through vendor support and failure analysis, that the low voltage power supplies (LVPSs) were the underlying issue to most of the historical and recent NUMAC failures.

The failures were initially attributed to the A10 card because voltage drift (usually to a higher voltage) of the LVPSs caused the A10 card to fail the NUMAC modules self - test function, which would then cause a self - test critical fault and led to unnecessary isolation signals. On some occasions, the A10 card was replaced, the NUMAC was retested, it passed, and the issue did not recur for some time. However, similar issues became more fre quent as years progressed. Since 2007, 25 such issues occurred and 2/3 of them have been subsequently determined to be LVPS related issues.

According to industry benchmarking performed by PSEG, PM of NUMAC LVPSs typically follows one of two approaches: 1 ) Proactive replacement at an interval of ~20 years, or 2) Regular monitoring of LVPS output voltages and adjustment/replacement as necessary.

From installation of these LVPSs in 1991, until 2008, PSEG had established measures to implement the first strategy.

However, in 2008, PSEG deactivated the replacement PM procedures while developing a performance centered maintenance (PCM) template for the components (70080081).

At that time, the action aligned with industry practices, and was justified by the fac t that the power supplies are redundant pieces of equipment.

The inspectors identified the following missed opportunities as they related to this review:

The inspectors determined that the 2008 PCM template review by PSEG was not adequate. The inspectors noted that the PCM template recommended actions to monitor critical parameters.

The stations approach to meeting this suggestion was to rely on the NUMAC microprocessors built-in self-test features that are meant to monitor various parameters and alarm when an issue develops.

However, until the underlying issue with LVPS voltage drift was identified in 2020, the station was not aware that LVPS voltage was a critical parameter that could cause failures without a prior alarm and therefore should have been monitored independently. Since this discovery, PSEG has proceduralized actions to regularly monitor LVPS voltages.

The inspectors noted that, in 2018, PSEG had a missed opportunity to correct the PM strategy deficiency.

During a vendor manual update review (70202000), PSEG self -

identified that regular monitoring of LVPSs should be performed to align with the vendor technical manuals maintenance recommendations. NOTF 20801250 was written to drive procedure updates for each of the NUMAC LDMs to include LVPS voltage monitoring. However, the procedure revisions were not completed because they were not identified as corrective actions associated with the NUMAC failures, and were therefore not prioritized, or completed until the station ultimately concluded that LVPS voltage drift was causing the issues.

The PM changes were proceduralized in November of 2020.

The inspectors evaluated these issues in accordance with the IMC 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues, and determined the issues did not constitute performance deficiencies or were of minor significance. As a result, these issues were not subject to enforcement action in accordance with the NRCs enforcement policy.

PSEG entered these observations into their corrective action program as NOTFs

EXIT MEETINGS AND DEBRIEFS

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

On October 19, 2021, the inspector s presented the integrated inspection results to Mr. Steve Poorman, Hope Creek Plant Manager, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.12 Corrective Action 20881718

Documents 20876449

Resulting from

Inspection

71111.15 Corrective Action 20879324

Documents 20879376

20879324

20879376

Corrective Action 20882398

Documents