Text

– Biennial Problem Identification and Resolution Inspection Report 05000382/2024010
	ML24331A244
Person / Time
Site:	Waterford
Issue date:	12/06/2024
From:	Brian Correll; NRC/RGN-IV/DORS/PBD
To:	Sullivan J; Entergy Operations
References
	IR 2024010
	Download: ML24331A244 (1)
	v • d • e

December 6, 2024

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000382/2024010

Dear Joseph Sullivan:

On October 24, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Waterford Steam Electric Station, Unit 3 and discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

The NRC inspection team reviewed the stations problem identification and resolution program to confirm that the station was complying with NRC regulations and licensee standards. The team also evaluated the stations effectiveness in identifying, prioritizing, evaluating, and correcting problems, reviewed licensee audits and self-assessments, and its use of industry and NRC operating experience information. Based on the samples reviewed, the team determined that your staff's performance in each of these areas adequately supported nuclear safety.

However, the team noted some challenges in the area of evaluation and prioritization of issues, and in the area of effectiveness of corrective actions. Specifically, the team noted examples where your staff missed opportunities to thoroughly evaluate issues to ensure the resolutions of each issue addressed the causes, and examples were the corrective actions for conditions adverse to quality were not implemented in a timely manner. The results of these evaluations are in the enclosure.

Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, the team found no evidence of challenges to your organizations safety-conscious work environment. Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.

Four findings of very low safety significance (Green) are documented in this report. Three 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Waterford Steam Electric Station, Unit 3.

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 IV; and the NRC Resident Inspector at Waterford Steam Electric Station, Unit 3.

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, Brian K. Correll, Inspection Programs & Assessment Team Leader Division of Operating Reactor Safety Docket No. 05000382 License No. NPF-38

Enclosure:

As stated

Inspection Report

Docket Number:

05000382

License Number:

NPF-38

Report Number:

05000382/2024010

Enterprise Identifier:

I-2024-010-0011

Licensee:

Entergy Operations, Inc.

Facility:

Waterford Steam Electric Station, Unit 3

Location:

Killona, LA 70057

Inspection Dates:

October 7, 2024, to October 24, 2024

Inspectors:

F. Ramirez Munoz, Senior Reactor Inspector

S. Hedger, Sr Emergency Preparedness Inspector

A. Sanchez, Senior Project Engineer

W. Schaup, Senior Project Engineer

Approved By:

Brian K. Correll, Team Leader

Inspection Programs & Assessment Team

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at Waterford Steam Electric Station, Unit 3, 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 Fully Implement the External Corrosion Program Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green FIN 05000382/2024010-01 Open/Closed

[H.1] -

Resources 71152B The inspectors identified a Green finding for the licensees failure to implement the site's external corrosion program in accordance with procedure UNT-006-032 Coating and Corrosion Program, revision 1. Specifically, the external corrosion program, which was created following a root cause evaluation for corrosion issues at the site in 2015, has, since 2018, been partially supported and has amassed a substantial backlog, estimated of more than 2500 items, with some dating back to 2011. It also includes approximately 200 items designated as severe that are associated with safety-related components.

Failure to Take Timely Corrective Action for an Adverse Condition Affecting Control Room Toxic Gas Detection Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000382/2024010-04 Open/Closed

[P.2] -

Evaluation 71152B The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI,

Corrective Actions, for the licensees failure to take timely corrective action for an adverse condition. Specifically, the licensee failed to address a vulnerability in their broad range gas monitors that could result in significant time delays in indication and protective action from control room staff during a hydrogen chloride release at the site.

Inadequate Work Instructions Result in Containment Spray Pump Breaker Failure Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000382/2024010-02 Open/Closed None (NPP)71152B The inspectors are documenting a self-revealed finding of very low safety significance (Green)and an associated non-citied violation of Technical Specification 6.8.1.a, "Instructions,

Procedures, and Drawings," for the licensee's failure to perform maintenance that can affect performance of safety-related equipment in accordance with written procedures appropriate to the circumstances. Specifically, station procedure ME-004-115, 4.16/6.9 kV Magne-Blast Breaker Overhaul, revision 6, was not appropriate to the circumstances because the procedure failed to contain instructions to ensure the containment spray pump breakers check nut associated with the opening spring was tightened to prevent mechanical binding of the trip latch and trip latch roller. The binding prevented the 'A' containment spray pump supply breaker from closing and operators had to declare the 'A' containment spray pump inoperable.

Inadequate Preventative Maintenance Strategy Results in Failure of Thermal Relief Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000382/2024010-03 Open/Closed None (NPP)71152B The inspectors are documenting a self-revealed finding of very low safety significance (Green)and an associated non-citied violation of Technical Specification 6.8.1.a, "Instructions,

Procedures, and Drawings," for the licensee's failure to implement a preventative maintenance schedule developed to specify inspection or replacement of parts that have a specific lifetime. Specifically, the licensee's preventative maintenance schedule failed to detect degradation of the main feed isolation valves thermal relief valve during preventative maintenance tasks, or to have tasks that would replace the thermal relief valves before a failure of the valves occurred.

Additional Tracking Items

None.

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.

OTHER ACTIVITIES - BASELINE

71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)

Main article: IP 71152

(1) The inspectors performed a biennial assessment of the licensees corrective action program, use of operating experience, self-assessments and audits, and safety-conscious work environment.

Corrective Action Program Effectiveness: The inspectors assessed the corrective action programs effectiveness in identifying, prioritizing, evaluating, and correcting problems. The inspection team evaluated the stations compliance with NRC regulations and licensee standards like procedures for implementing the corrective action program. The inspectors sampled approximately 280 condition reports and their associated cause evaluations, as applicable. The inspectors also conducted a five-year review of the feedwater and feedwater control systems.

The inspectors also reviewed corrective actions associated with enforcement discretion items documented in the inspection procedure 95001 Supplemental Inspection Report 05000382/2023040 (ML23192A764), dated July 12, 2023.

In addition, the inspectors reviewed three of the causal products associated with the inspection procedure 95001 inspection (condition reports (CR) CR-WF3-2022-03999, CR-WF3-2022-06367, and CR-WF3-2022-07836) to verify if actions planned at the end of the previous inspection were completed or were scheduled based on reasonable justifications. Additional corrective actions had been completed, and effectiveness review actions are continuing on schedule.

The corrective actions for the following NCVs and findings were evaluated as part of the assessment:

NCV 05000382/2023010-01; NCV 05000382/2023401-01; NCV 05000382/2024004-02; NCV 05000382/2023401-03; NCV 05000382/2023401-02; NCV 05000382/2022013-01; NOV 05000382/2022091-01; NCV 05000382/2022013-03; NCV 05000382/2022013-02; NCV 05000382/2022002-02; NCV 05000382/2022002-01; NCV 05000382/2022002-06; NCV 05000382/2022002-05; NCV 05000382/2022002-08; NCV 05000382/2022002-07; NOV 05000382/2022501-01; and the minor violations tracked by the licensee from inspection report 05000382/2022010.

Operating Experience, Self-Assessments and Audits: The inspectors assessed the effectiveness of the stations processes for use of operating experience, audits and self-assessments. The sample included industry operating experience communications like 10 CFR Part 21 notifications and other vendor correspondence, NRC generic communications, publications from industry groups, and site evaluations. The sample also included reviews of licensee self-assessments and internal audits.

Safety-Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety-conscious work environment. The team interviewed 43 individuals, observed interactions between licensee employees and management during routine meetings, interviewed the employee concerns program specialist and reviewed employee concerns files.

`

INSPECTION RESULTS

Assessment 71152B Corrective Action Program Effectiveness Based on the samples reviewed, the team determined that the licensee's corrective action program complied with regulatory requirements and self-imposed standards. The licensee's performance in each of the areas of Problem Identification, Problem Prioritization and Evaluation, and Corrective Actions adequately supported nuclear safety.

Problem Identification The inspectors found that the licensee was generally identifying and documenting problems at an appropriately low threshold that supported nuclear safety. Conditions that required generation of a condition report had been identified and entered appropriately into the corrective action program.

However, the team identified one example where the licensee didn't use corrective action program data and other applicable insights (such as system health reports) to identify low level trends with equipment performance. Specifically, in reviewing CR-WF3-2021-04956, the inspectors noted that the licensee's preventative maintenance schedule failed to detect degradation of the main feed isolation valves thermal relief valves during preventative maintenance tasks or to have tasks that would replace the thermal relief valves before a failure of the valves occurred. This report documents a self-revealed non-cited violation for this issue.

Problem Prioritization and Evaluation Overall, the team found that the licensee was appropriately prioritizing and evaluating issues to support nuclear safety. Of the samples reviewed, the team generally found that the licensee correctly characterized condition reports as to whether it represented a condition adverse to quality, and then prioritized the evaluation and corrective actions in accordance with the program guidance.

However, the team identified four examples where the licensee missed an opportunity to thoroughly evaluate an issue to ensure the resolutions of each issue addressed the causes of it.

CR-WF3-2023-18180 - In January 2024, the site received a Severity Level IV non-cited violation (05000382/2023010-01) for failure to verify the design adequacy of the alternate source term dose consequence analysis and the failure to provide complete and accurate information to the NRC. The inspectors noted during the review of the corrective actions performed to date that the licensee did not prioritize addressing the failure to provide complete and accurate information to the NRC and focused much of their evaluation on resolving the adequacy of the design of the alternate source term dose consequence analysis. As a result, the licensee failed to have corrective actions to address that part of the violation. This report documents an Observation for this issue.

CR-WF3-2022-06858 - In October 2022, control room operators misdiagnosed the failure of the air ejector particulate, iodine, and gas radiation monitor and took action to install alternate sampling on the main condenser wide range gas monitor which rendered main condenser wide range gas monitor inoperable for 20 minutes. The condition report associated with the issue, subsequent training, and evaluations focused on maintaining configuration control of equipment in general, and the consequences of working on the wrong equipment and resulting technical specification entries. However, the inspectors noted that the licensee did not address that the main condenser wide range gas monitor was needed for the dose assessment capability required for emergency preparedness. This inspection report documents a minor violation of NRC requirements associated with this issue.

CR-WF3-2024-01503 - In March 2024, three in-service steam bypass control valves MS-319A, MS-319B, and MS-319C closed unexpectedly during a reactor startup due to the steam bypass control valves permissive being removed, and resulted in pressurizer pressure rising to 2285 psia. All three steam bypass valves remained closed for approximately 5 seconds then reopened when the permissive became active. This same event occurred again 5 minutes later at which point all three steam bypass control valves permissive switches were taken to manual. This event resulted in an unplanned LCO entry due to the two momentary raises in reactor pressure above 2275psi. The inspectors noted that the evaluation and disposition of this issue failed to include the causes for the unexpected closure and an assessment on whether there was an equipment issue or an operator training issue. Instead, the licensee disposition of the issue referred to existing guidance for operating the steam bypass control system valves. The licensee generated CR-WF3-2024-05069 to document this issue.

CR-WF3-2022-05993 - Inspectors reviewed non-cited violation 05000382/2022002-02 where radiation workers received an uptake of airborne radioactive material in April 2022, because the licensee failed to appropriately implement aspects of EN-RP-105, Radiological Work Permits (RWP), to maintain doses ALARA. This non-cited violation had an Operating Experience (P.5) cross-cutting aspect assigned to it because in 2015, the licensee was performing similar activities and a radiation worker received contamination on the face. The licensee wrote CR-WF3-2022-05993 to document the violation and closed it to CR-WF3-2022-02805, which had been written when the actual uptake event occurred. However, when addressing CR-WF3-2022-02805, the resolution of the issue failed to address the P.5 cross-cutting aspect of the violation by not evaluating why the previous contamination event had been missed.

Further, the inspectors noted that radiation worker behaviors were not evaluated in their causal product. The licensee generated CR-WF3-2024-05671 to address this issue.

Corrective Actions Overall, the team concluded that the licensee's corrective actions supported nuclear safety.

Specifically, Waterford developed effective corrective action for the problems evaluated in the corrective action program. The inspectors also determined that the licensee generally implemented these corrective actions in a timely manner commensurate with their safety significance.

However, the inspectors identified four examples of issues were the corrective actions for conditions adverse to quality were not implemented in a timely manner.

CR-WF3-2021-06586 - The inspectors reviewed an adverse condition identified in November 2021, for an issue with the detection of hydrogen chloride concentrations in the licensees broad range gas monitors that resulted in delayed detection of hydrogen chloride for up to 80 minutes. Credit is being taken for being alerted of a toxic gas release in the area through the St. Charles Parish Industrial Hotline. Based on a review of the licensees design basis, it is assumed that this phone notification would take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to receive. If hydrogen chloride levels in the vicinity of the control room are at toxic levels (immediately dangerous to life and health), the analysis assumes that operators would have permanent health effects with 30 minutes of exposure. The credited basis document for the licensees analysis assumes protective measures are provided within 2 minutes of detection. Reliance on the delayed detection, or the phone call, will not provide for the assumed operator protection. The licensee did not recognize the assumed detection method timelines would not provide for worker protection in a timeline consistent with the design basis and has not provided timely corrective actions since the condition was first identified. A 10 CFR 50, Appendix B, Criterion XVI Green non-cited violation is documented in this inspection report for this issue.

CR-WF3-2019-04811 - In May 2019, a generator trip and subsequent reactor cutback occurred. The steam generators experienced a level deviation that placed feedwater control system in manual and did not result in the expected reactor trip override that repositions the feedwater regulating valves. To correct the identified condition, a modification was implemented to establish a level deviation setpoint that would not shift feedwater control to manual following reactor power cutbacks. In 2020, CR-WF3-2020-05623 documented that the prescribed deviation could not be achieved, and in 2021, CR-WF3-2021-00483 documented that the margin gained during the modification was less than originally evaluated. The inspectors noted that the physical cause of the 2019 event still existed at the site on March 21, 2024, during a reactor trip and that the licensee hadnt corrected the condition. This report documents an Observation for this issue.

CR-WF3-2021-05760 - The inspectors reviewed non-cited violation 05000382/2022002-08 where the licensee failed to appropriately verify the adequacy of the shield building ventilation design and as a result was inoperable for longer than allowed by Technical Specifications. The inspectors noted that the licensee had instituted interim corrective actions to restore compliance and that the permanent corrective actions were completed in 2024. The inspectors noted the issue was discovered in 2021, the non-cited violation was documented in 2022, and the corrective actions were completed in 2024. This was determined to be unnecessarily long, especially considering that the action was to correct the design issue. The licensee generated CR-WF3-2024-05605 to document this issue.

CR-WF3-2015-00947 - The inspectors reviewed Waterford's corrosion program, which was created in 2015, to identify, prioritize and resolve external corrosion issues in plant equipment, following a root cause evaluation for corrosion issues at the site in 2015. The inspectors reviewed the status of the corrosion program and noted that the program is no longer being implemented and both staffing and funding for the program have been abandoned. This has resulted in a backlog of at least 2500 items, which date back to 2011, of which, approximately 200 are listed as severe. The inspectors concluded that the actions to resolve and address corroded equipment has not been timely. A finding is documented for this issue in this inspection report.

Assessment 71152B Use of Industry and NRC Operating Experience The team reviewed a variety of sources of operating experience including part 21 notifications and other vendor correspondence, and NRC generic communications. The team determined that, overall, the licensee is adequately screening and addressing issues identified through operations experience that apply to the station, and this information is being evaluated in a timely manner once it is received.

Assessment 71152B Audits and Self-Assessments The inspectors reviewed a sample of Waterford's self-assessments to assess whether performance trends were regularly identified and effectively addressed. The inspectors also reviewed audit reports to assess the effectiveness of assessments in specific areas. Overall, the inspectors concluded that the licensee had an adequate departmental self-assessment and audit process.

However, the inspectors identified one example where the recommendations from a self-assessment were not considered, and corrective actions were not planned. Specifically, the inspectors reviewed a 2022 self-assessment dated March 16, 2022, which was associated with the corrosion program and was in response to a NIOS audit. It was not apparent that the results were acted on, and did not address the ailing program. A finding associated with this issue is documented in this inspection report.

Assessment 71152B Safety-Conscious Work Environment The team conducted safety-conscious work environment interviews with 43 employees from six different disciplines that included mechanical maintenance, instrumentation and controls, electrical maintenance, operations, security, and chemistry. The purpose of these interviews was:

(1) to evaluate the willingness of the licensee staff to raise nuclear safety issues, either by initiating a condition report or by another method,

(2) to evaluate the perceived effectiveness of the corrective action program at resolving identified problems, and

(3) to evaluate the licensee's safety-conscious work environment (SCWE). The team also observed interactions between employees during routine department performance improvement coordinators (DPIC) prescreening meetings, performance improvement review group meetings, and operations focus meetings.

The team found that the licensee had a safety-conscious work environment where individuals felt free to raise concerns without fear of retaliation. Most expressed positive experiences after raising issues to their supervisors and after documenting issues in condition reports, and all individuals indicated that they would not hesitate to raise safety concerns through at least one of the several means available at the station.

Based on feedback from these interviews, all the groups stated that nuclear safety issues are addressed promptly and in a timely manner. However, some of the groups expressed frustration with the length of time to resolve lower tier issues, and that even when its engrained in the staff to write a condition report when an issue is found, it sometimes requires multiple condition reports being written before a lower tier issue is addressed. Most groups knew that the Employee Concerns Program and the Hotline were avenues where they could submit anonymous concerns if they had to but were not aware of the specific details on how to submit one. Some groups brought up morale concerns, which makes it important for management in all departments to remain a strong and reliable avenue where employees feel free to bring up safety concerns. When these avenues start to erode, it could eventually impact licensee staff's willingness to bring up concerns using the management avenue and negatively impact employees confidence in the corrective action program.

The team looked at the stations Employee Concerns Program. The team also interviewed the Employee Concerns Program Specialist and reviewed a sample of case files that were opened during the inspection period. The team noted that in general, the focus groups expressed willingness to use the Employee Concerns Program. Overall, the team did not identify any concerns with the program.

Failure to Fully Implement the External Corrosion Program Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green FIN 05000382/2024010-01 Open/Closed

[H.1] -

Resources 71152B The inspectors identified a Green finding for the licensees failure to implement the site's external corrosion program in accordance with procedure UNT-006-032 Coating and Corrosion Program, revision 1. Specifically, the external corrosion program, which was created following a root cause evaluation for corrosion issues at the site in 2015, has, since 2018, been partially supported and has amassed a substantial backlog, estimated of more than 2500 items, with some dating back to 2011. It also includes approximately 200 items designated as severe that are associated with safety-related components.

Description:

In February 2015, performance improvement management wrote CR-WF3-2015-0947 to document an on-going issue with identification, evaluation and resolution of external corrosion, as exemplified by 368 unique condition reports on external corrosion (since 2010),and three 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, violations issued due to the stations inability to identify and correct corrosion issues. The condition report was classified as a category A significance which required a root cause evaluation and the issuance of corrective actions to prevent repetition (CAPR). The root cause evaluation was that the station maintenance and production process standards do not adequately identify, prioritize, monitor, and promptly resolve degradation at the necessary level to promote and support a low threshold for identification of external corrosion by station personnel. The main contributing cause was the site leadership team behaviors were not effective in correlating the increase in external corrosion degradation to weaknesses in process and personnel behaviors. This issue was supplemented with the lack of an effective preventative maintenance strategy to proactively paint/preserve and/or replace equipment susceptible to external corrosion. More specifically, preventative maintenance activities lack scope, budget and resources, and quantitative criteria for satisfactory completion of activities.

The station identified the following CAPRs to prevent the issue from recurring: 1) using an external subject matter expert, develop a repeatable and reproducible standard for the identification of external corrosion, and 2) develop and implement an external corrosion program. Further, the program included, in part, the following: 1) establishment of an external corrosion program owner, and a maintenance external corrosion owner, 2) establish guidance on identification, prioritization, monitoring and repairing degradation of plant equipment, 3) requirement to monitor identified corrosion items in the backlog greater than one year for change in condition, 4) perform an assessment every three years as a method to monitor program and plant performance in relation to external corrosion, and 5) guidance on development of long term funding. The licensee successfully established the framework for an external corrosion control via procedure UNT-006-032, Coating and Corrosion Program, revision 0, in March 2016.

While onsite during the problem identification and resolution inspection, the inspectors requested plant walkdowns of the essential chill water system and the ultimate heat sink (wet and dry cooling towers). The walkdown revealed concerns with the amount of corrosion. More specifically, the corrosion items were identified but had been waiting to be corrected. Through review of CR-WF3-2015-0947, discussions with the corrosion marshal (past and present) and the code engineer, a review of the backlog of corrosion issues that need to be remedied, a review of the last program self-assessment, and program procedures, the inspectors determined that the external corrosion program has not been properly supported by site management since approximately 2018. The program and site personnel continue to identify issues, but those issues have languished due, in part, to inconsistent site leadership financial support and oversight. The team noted that the program would get funding, then funding would be pulled. This resulted in inconsistent and fragmented progress and has made it difficult to hire contractors to support repairs if they might get laid off in the middle of the contract. The lack of site leadership oversight has often left the program vulnerable without a champion to ensure that the program continues to function reliably to identify and correct external corrosion issues.

To better understand the number and type of external corrosion issues, the team requested a list of issues (CRs) and prioritization of those issues. The corrosion marshal and the system engineer could not fully complete the request. There are two tracking systems, one involving an Excel spreadsheet, and a second in Maximo. The staff expressed frustration due to the confusing process and significant amount of backlogged items. From the information provided, the team determined that there are more than 2500 items being tracked, some dating back to 2011, and approximately 200 are classified as severe (the highest priority according to the corrosion program) and associated with safety-related systems.

A review of the last program self-assessment in 2022, revealed that the external corrosion program was in declining health, with no corrosion marshal position staffed since 2018, preventative maintenance walkdowns not being performed according to the program, code programs engineer not even aware of the programs existence, and lack of qualified personnel to implement the program. The assessment further concludes that it suffers from lack of site leadership oversight. The team also determined that the licensee was presented an opportunity in the form of a self-assessment with a strong message, but no condition reports were initiated to document the findings to be resolved.

The team also identified that the current set of workers involved in implementing the program to resolve issues (including maintenance), is severely understaffed and requires assistance from the rest of the site to assess the current number of external corrosion issues (multiple confusing tracking mechanisms), and creating a systematic plan to address these issues in a timely manner.

Corrective Actions: The issue was placed into the corrective action program.

Corrective Action References: CR-WF3-2024-05211

Performance Assessment:

Performance Deficiency: The failure to implement corrosion program in accordance with procedure UNT-006-032 Coating and Corrosion Program, revision 1, was a performance deficiency. Specifically, the external corrosion program, which was created following a root cause evaluation for corrosion issues at Waterford in 2015, has, since 2018, been partially supported and has amassed a substantial backlog, estimated at more than 2500 items, with some dating back to 2011, and includes approximately 200 items designated as severe and associated with safety-related components.

Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, not implementing timely corrective action to resolve corrosion issues could affect initiating and mitigating systems and components and lead to more significant issues.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Exhibit 2, Mitigating Systems Screening Questions. The finding was determined to have very low safety significance (Green) because it is not a deficiency affecting the design or qualification of a mitigating SSC; did not represent a degraded condition that resulted in a loss of probabilistic risk analysis (PRA) function of a single train of a technical specification (TS) system; did not represent a degraded condition that resulted in a loss of PRA function of one train of a multi-train TS system; did not represent a degraded condition that resulted in the loss of the PRA function of two separate TS systems; did not represent a degraded condition that resulted in the loss of the PRA function as defined in the Plant Risk Information e-Book (PRIB) or licensees PRA; and did not represent a degraded condition that resulted in the loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensees maintenance rule program.

Cross-Cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.

Specifically, licensee management, and site leadership, failed to adequately oversee, staff and fund the external corrosion program which has led to extensive backlog of issues that are not being addressed in a timely manner.

Enforcement:

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

Failure to Take Timely Corrective Action for an Adverse Condition Affecting Control Room Toxic Gas Detection Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000382/2024010-04 Open/Closed

[P.2] -

Evaluation 71152B The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take timely corrective action for an adverse condition. Specifically, the licensee failed to address a vulnerability in their Broad Range Gas Monitors that could result in significant time delays in indication and protective action from control room staff during a hydrogen chloride release near the site.

Description:

The inspectors reviewed CR-WF3-2021-06586 (November 22, 2021) dealing with an issue identified with detection of hydrogen chloride (HCl) concentrations in the licensees Broad Range Gas Monitors (BRGMs). HCl, one of a number of toxic constituents monitored by the BRGMs, adheres to the materials within the sample line. This adherence causes time delays in detecting HCl levels of concern that exceed the time assumptions in the licensing basis. Data available to the licensee indicated that measurement of any given HCl level can be delayed by up to 80 minutes. With this existing condition, the licensee is relying on notification of a toxic gas release in the area through the St. Charles Parish Industrial Hotline to compensate for the condition.

The inspectors reviewed the details of the condition report to determine what corrective actions have been completed or planned. Based on what was provided upon inspection, no corrective actions to address the condition have been completed. This condition was categorized as a Long Term Corrective Action, indicating that an extended time was justified in taking corrective action. The Long Term Corrective Action Classification Form (dated December 16, 2021) indicated that a modification or design change was required to address this condition. However, no modification or design change documentation was available for review. The condition report or its provided attachments did not explain any circumstances with the modification that would justify scheduling action at a later time. Corrective action number CR-WF3-2021-06586-00003 indicated that corrective action was scheduled for December 2024 but provided no explanation of what the planned corrective actions would be.

Review of the licensing basis documentation provided more insight into what time assumptions were made regarding detection and response. Waterford SES Unit 3 Final Safety Analysis Report, revision 320, discusses the design evaluation for toxic gas protection. Sections 2.2.3.3 and 6.4.4.2 describe the toxic chemical design basis, analysis, and designed protective measures. Review of these shows the following:

Section 2.2.3.3 indicates that the toxic chemical analysis was developed using Regulatory Guide 1.78, revision 0 (June 1974).

Sections 2.2.3.3.5 and 2.2.3.3.6 indicate that the analysis assumes that major accidental releases of toxic chemicals will result in site notification via the St. Charles Parish Industrial Hotline within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Section 2.2.3.7c describes action taken as a result of the Broad Range Gas Detection System. If toxic gas concentrations equal or exceed its setting levels, the detector system sounds an alarm and automatically isolates the control room before toxic or Immediately Dangerous to Life or Health (IDLH) levels can be reached. The systems maximum response time for measured gases is 17 seconds or less.

Regulatory Guide 1.78, revision 0, explains expectations for assumed times for chemical detection and response. Section C.4 says for each chemical analyzed, the values of importance are the detection threshold and the maximum concentration that can be tolerated for two minutes without physical incapacitation of the average human. This maximum concentration is the toxicity limit, or as clarified in revision 1 of the same guide (section B, December 2001), the IDLH limit. Thirty minutes exposure to a chemical at its IDLH limit can result in delayed or immediate permanent health effects if no protective action is afforded.

Further it clarifies that adequate margin for control room operator safety is assumed as long as protective measures are used within 2 minutes of chemical detection. It is noted that exposure to IDLH limits for this amount of time is acknowledged in the licensees design basis, per Final Safety Analysis Report (FSAR) section 2.2.3.3.6. Based on review of the condition report, and discussion with licensee staff, there was not awareness of time assumptions that were part of the supporting analyses. Therefore, they were not taken into account in decision-making about corrective actions.

For the as-found condition during a HCl release, protective measures for control room operators could not be ensured, consistent with the design basis and its supporting documentation. Specifically, a HCl release resulting in IDLH levels in the control room vicinity would not be detected by the BRGMs for approximately 80 minutes, and no assumed detection by the St. Charles Parish Industrial Hotline for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months based on their analysis. This would not provide for the protective measures to detectable HCl levels within two minutes when it should be detected, let alone protect against control room operators exposure exceeding 30 minutes.

Corrective Actions: The licensee entered these issues into the corrective action program.

Corrective Action References: Condition reports CR-WF3-2021-06586 and CR-WF3-2024-05209

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure to provide timely corrective action to an adverse condition is a performance deficiency within the licensees ability to foresee and correct. Specifically, the licensee failed to meet procedure EN-LI-102, section 5.7, Corrective Actions, step 5.7.2, Corrective Action Initiation, which states that corrective action content should be specific, measurable, actionable, reasonable, and timely.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The licensees ability to protect the public from accidents and events is adversely affected when the control room operators health is negatively impacted.

Significance: The inspectors assessed the significance of the finding using IMC 0609 appendix A, The Significance Determination Process (SDP) for Findings At-Power, exhibit 3, section D.2 (issue date November 30, 2020); a detailed risk evaluation was required to determine the safety significance of the finding. The analyst assumed that the condition resulting from the performance deficiency potentially affected the capability of the control room ventilation system to provide adequate protection for control room operators in an event involving hydrogen chloride toxic chemical release. The analyst reviewed licensee Calculation EC-S97-020, Toxic Chemical Analysis to Assess Control Room Habitability, revision 1, and determined that the best available information regarding an estimation of the frequency of a toxic gas release event affecting the station involving the hydrogen chloride chemical shows that such an event could be estimated to occur with a frequency of 4.1E-7 per year. Accordingly, the analyst concluded that any increase in average annual core damage frequency attributable to the condition resulting from the performance deficiency would be less than 1.0E-6 per year. The analyst also determined that there was no appreciable increase in large early release frequency attributable to this condition. Therefore, the finding was determined to be of very low safety significance (Green).

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the evaluation of the issue did not recognize the time assumptions in the licensing basis and related documentation. Had these assumptions been understood at the time, it would have affected the significance of the condition, resulting in a different approach to resolving the issue.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly corrected.

Contrary to the above, from November 22, 2021, to October 24, 2024, Waterford 3 failed to establish measures to assure conditions adverse to quality are promptly corrected.

Specifically, as of the exit meeting date on October 24, 2024, the licensee failed to correct an adverse condition with their BRGMs within a reasonable time to provide control room operator protection against a hydrogen chloride release.

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

Inadequate Work Instructions Result in Containment Spray Pump Breaker Failure Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000382/2024010-02 Open/Closed None (NPP)71152B The inspectors are documenting a self-revealed finding of very low safety significance (Green) and an associated non-citied violation of Technical Specification 6.8.1.a, "Instructions, Procedures, and Drawings," for the licensee's failure to perform maintenance that can affect performance of safety-related equipment in accordance with written procedures appropriate to the circumstances. Specifically, station procedure ME-004-115, 4.16/6.9 kV Magne-Blast Breaker Overhaul, revision 6, was not appropriate to the circumstances, because the procedure failed to contain instructions to ensure the containment spray pump breakers check nut associated with the opening spring was tightened to prevent mechanical binding of the trip latch and trip latch roller. The binding prevented the 'A' containment spray pump supply breaker from closing and operators had to declare the 'A' containment spray pump inoperable.

Description:

On September 29, 2022, 'A' containment spray pump breaker did not close while attempting to start the pump following maintenance under work order (WO) 517244. The first start attempt was in accordance with station procedure OP-903-094, ESFAS Subgroup Relay Test, section 7.29. The test pushbutton was depressed, the pump did not start and annunciator M0104, Containment Spray Pump A Unavailable, was received. Once the test pushbutton was released, the annunciator cleared. The control switch for the pump was taken to start and the pump did not start. Since the pump could not start, the Operations staff declared the 'A' containment spray pump inoperable. The licensee documented the event in CR-WF3-2022-06818.

During the initial investigation of the breaker, it was noted that the breakers trip latch and the roller bearing did not have enough clearance from each other. This would cause the trip latch to occasionally catch on the bearing during test cycles causing the breaker to immediately trip when taken to close. The breaker was overhauled in February 2021, under WO 52790255 and passed the clearance checks.

The licensee performed troubleshooting under WO 54034973. The cause of the breaker failing to open was found to be mechanical binding between the trip latch and the trip latch roller on the breaker. The clearance between the two parts were found to be out of tolerance because of the stop plate shifting. The stop plate is fastened in place by a check nut. The check nut was found loosened in the field.

The licensee performed an equipment failure evaluation that determined that the casual factor for the mechanical binding of trip latch and trip latch roller was a misadjusted spring plate due to a loose check nut. This was the result of inadequate work instructions in station procedure ME-004-115, 4.16/6.9 kV Magne-Blast Breaker Overhaul, revision 6. The procedure did not have a step to check the tightness of the check nut after adjusting the opening spring for trip latch and roller tolerance.

The inspectors reviewed the applicable condition reports, causal products and documents and identified the following:

The stations technical specifications section 6.8.1.a. require that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978.

NRC Regulatory Guide 1.33, revision 2, appendix A, Section 9.a requires that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures appropriate to the circumstances.

The inspectors determined that the maintenance performed under WO 52790255 that used station procedure ME-004-115, was not appropriate to the circumstance because the station procedure did not have a step to check the tightness of the check nut to ensure that the breaker would close to provide power to the pump to perform its safety function.

Corrective Actions: The failed breaker was replaced with a spare breaker and the 'A' containment spray pump was tested satisfactorily and declared operable. Station procedure ME-004-115 was revised with additional instructions to check the tightness of the check nut.

Corrective Action References: Condition reports CR-WF3-2022-06818 and CR-WF3-2024-05207

Performance Assessment:

Performance Deficiency: Regulatory Guide 1.33, section 9.a requires that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures appropriate to the circumstances. The inspectors determined that the failure to ensure that station procedure ME-004-115, 4.16/6.9 kV Magne-Blast Breaker Overhaul, contained instructions to ensure the check nut associated with the opening spring was tightened to prevent mechanical binding of the trip latch and trip latch roller was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the check nut became loose resulting in the trip latch and trip latch roller binding which immediately tripped the breaker for 'A' containment spray pump.

With the breaker not able to be closed the containment spray pump had to be declared inoperable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the issue using Exhibit 3, Barrier Integrity Screening Questions, section C, Reactor Containment and answered "Does the finding represent an actual open pathway in the physical integrity of reactor containment, the failure of containment isolation system (logic and instrumentation), failure of containment pressure control equipment (including SSCs credited for compliance with Order EA-13-109), failure of containment heat removal components, or failure of the plants severe accident mitigation features (AP1000)?"

as YES and proceeded to IMC 0609, Appendix H.

A detailed risk evaluation was required to determine the safety significance of the finding. For the detailed risk evaluation, the analyst assumed an exposure time of 90 days, assuming the breaker would not have functioned since the last quarterly surveillance run. The analyst set basic event CSS-MDP-FS-A, Containment Spray/Recirculation Motor Driven Pump Train A Fails to Start, to TRUE in the Waterford SPAR model, version 8.81, run on SAPHIRE, version 8.2.11, to estimate an increase of 4.0E-8/year in core damage frequency from internal events. External events were assumed to not be appreciable contributors to the significance.

The analyst referenced Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, and classified the finding as a Type A finding since it affected both core damage frequency (CDF) and large early release frequency (LERF). Using Figure 6.1, Road Map for LERF-based Risk Significance Evaluation for Type A Findings at-Power, the inspectors determined that since the increase in CDF was less than 1.0E-7/year, use of the increase in CDF was appropriate to characterize the significance of the finding.

Therefore, the inspectors characterized the significance of this finding to be very low (Green).

The dominant sequence involved stuck open relief valve initiators which were mitigated by the high-pressure injection and recirculation.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Technical Specification section 6.8.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978. NRC Regulatory Guide 1.33, revision 2, appendix A, section 9 addresses Procedures for performing Maintenance and section 9.a requires that maintenance that can affect the performance of safety-related equipment be performed in accordance with written procedures appropriate to the circumstances.

Contrary to the above, from February 3, 2021, until October 5, 2023, the licensee failed to perform maintenance that can affect performance of safety-related equipment in accordance with written procedures appropriate to the circumstances. Specifically, procedure ME-004-115, 4.16/6.9 kV Magne-Blast Breaker Overhaul, revision 6 was not appropriate to the circumstances, because the procedure failed to contain instructions to ensure the check nut associated with the opening spring was tightened to prevent mechanical binding of the trip latch and trip latch roller.

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

Inadequate Preventative Maintenance Strategy Results in Failure of Thermal Relief Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000382/2024010-03 Open/Closed None (NPP)71152B The inspectors are documenting a self-revealed finding of very low safety significance (Green) and an associated non-citied violation of Technical Specification 6.8.1.a, "Instructions, Procedures, and Drawings," for the licensee's failure to implement a preventative maintenance schedule developed to specify inspection or replacement of parts that have a specific lifetime. Specifically, the licensee's preventative maintenance schedule failed to detect degradation of the main feed isolation valves thermal relief valves during preventative maintenance tasks or to have tasks that would replace the thermal relief valves before a failure of the valves occurred.

Description:

On September 8, 2021, the reactor auxiliary building watch operator reported main feed isolation valve (MFIV) No.1 nitrogen accumulator 'A' pressure as 5930 psig and main feed isolation valve No. 2 nitrogen accumulator 'B' pressure as 5939 psig. The control room operators identified that the pressures were greater than the operability limit of 5900 psig as specified in OP-100-014, Technical Specification and Technical Requirement Compliance, and declared the main feed isolation valves inoperable. The licensee documented the event in CR-WF3-2021-4947.

On September 9, 2021, the licensee wrote CR-WF3-2021-4956, that stated the nitrogen pressure on one of the accumulators on each of the MFIV was found to be above the operability limit of 5900 psig specified in station procedure OP-100-014. It is expected that there will be significant MFIV accumulator pressure increases as a result of power ascension during plant startup due to rapidly increasing system temperatures. However, for the maximum as-found accumulator pressures reported in CR-WF3-2021-4947 to occur, the corresponding thermal relief valves (FW-18410A and FW-18411B) would have to fail to lift based on the relief valves setpoint of 5800 psig +/- 100 psig. Therefore, it appears as though FW-18410A (MFIV 1 Side A (L Valve) Hydraulic Accumulator Thermal Relief) and FW-18411B (MFIV 2 Side B (L1 Valve) Hydraulic Accumulator Thermal Relief) did not lift when required to maintain pressure below a maximum of 5900 psig.

As part of the licensee's investigation the failed thermal relief valves were removed and sent for failure analysis. The analysis stated that a foreign substance layer was found on the pressure ports and within one of the valves. During manual operation of the poppet during disassembly of the valve, there was indication of binding or hesitation. It was suspected that the foreign substance created interference and increase friction to sliding, which could increase the apparent setpoint of the valve. Additionally, after having discussions with the vendor that performs the refurbishment of the MFIVs, it was discovered that the poppet assembly of the MFIV accumulator thermal relief valves are not replaced as part of the overall refurbishment and that only the manifold interface O-rings are replaced with respect to these thermal relief valves.

The licensee performed an equipment failure evaluation that determined an inadequate preventative maintenance strategy for the MFIVs that did not include a complete replacement of the accumulator thermal relief valves within the refurbishment process allowed for long term degradation of the hydraulic fluid to negatively impact the lift setpoint of the failed relief valves.

The inspectors reviewed the applicable condition reports, causal products and documents and identified the following:

The stations TS 6.8.1.a. require that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978.

NRC Regulatory Guide 1.33, revision 2, appendix A, section 9.b requires that preventative maintenance schedules be developed to specify inspection or replacement of parts that have a specific lifetime.

The inspectors determined that the licensee's preventative maintenance schedule for the main feed isolation valves failed to detect degradation of the thermal relief valves during preventative maintenance tasks or to have tasks that would replace the thermal relief valves before a failure of the valves occurred resulting in the main feed isolation valves being unable to perform their safety function.

Corrective Actions: The immediate corrective actions were to have maintenance personnel restore the accumulator pressures to within band.

Part equivalence evaluation 222817 was performed to put controls in place to ensure that that subject thermal relief valves (FW-18410A/B and FW-18411A/B) are always replaced with new valves as part of the MFIV Actuator refurbishments in the future. Completed March 16, 2023.

Part equivalence evaluation 222818 was performed to put controls in place to ensure that only new valves are used as spares for Cat ID 0001010017 to prevent recurrences when a refurbished spare was used to replace FW-18410A. Completed March 20, 2023.

Corrective Action References: Condition reports CR-WF3-2021-04956 and CR-WF3-2024-05208

Performance Assessment:

Performance Deficiency: NRC Regulatory Guide 1.33 requires that preventative maintenance schedules be developed to specify inspection or replacement of parts that have a specific lifetime. The inspectors determined that the licensee's preventative maintenance schedule failed to detect degradation of the thermal relief valves during preventative maintenance tasks or to have tasks that would replace the thermal relief valves before a failure of the valves occurred and was therefore a performance deficiency.

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 failed relief valves allowed nitrogen accumulator pressure to exceed station procedure OP-100-014 operability limits of 5900 psig. With nitrogen accumulator pressure above 5900 psig the main feed isolation valves had to be declared inoperable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating SSC, was not a degraded condition that represented a loss of the PRA function of a single train TS system for greater than its TS allowed outage time, was not a degraded condition that represented a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time, was not a degraded condition that represented a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , was not a degraded condition that represented a loss of a PRA system and/or function as defined in the PRIB or the licensees PRA (such as recovery of offsite power or the ability to feed and bleed) for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and was not a degraded condition that represented a loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensees maintenance rule program for greater than 3 days.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Technical Specification 6.8.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, revision 2, appendix A, February 1978. NRC Regulatory Guide 1.33, revision 2, appendix A, section 9, "Procedures for Performing Maintenance."

Part b of section 9, states, in part, that preventative maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime.

Contrary to the above, From March 29, 2018, until March 20, 2023, the licensee failed to implement preventative maintenance schedules developed to specify inspection or replacement of parts that have a specific lifetime. Specially, the main feed isolation valve preventative maintenance schedule failed to identify degradation or replace parts that have a specified lifetime long-term resulting in degradation of the thermal relief valves due to hydraulic fluid that negatively impacted the lift set point of the failed valves.

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

Minor Violation 71152B Failure to Maintain the Effectiveness of the Emergency Plan Minor Violation: Title 10 CFR 50.54(q)(2) requires a licensee shall maintain the effectiveness of the emergency plan that meets the requirements in appendix E to this part, and the planning standards of 50.47(b). Title 10 CFR 50.47(b)(9) says adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use. Contrary to this requirement, on October 1, 2022, the licensee failed to maintain the effectiveness of the emergency plan. Specifically, from 12:01 p.m. to 12:21 p.m., adequate methods and equipment for assessing and monitoring consequences of a radiological emergency condition were not in use. The main condenser wide range gas monitor (PRMIR0002), used for dose assessment purposes for some accident events, was inoperable without compensatory measures established.

Screening: The inspectors determined the performance deficiency was minor. Based on review of more-than-minor questions in IMC 0612, appendix B, informed by emergency preparedness significance determination process examples for 50.47(b)(9) performance deficiencies (IMC 0609, appendix B, Table 5.9-1), this issue screens to minor.

Enforcement:

The licensee has taken actions to restore compliance as documented in CR-WF3-2022-06858. This failure to comply with 10 CFR 50.54(q)(2) constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

Observation: Missed Opportunities to Address the Reactor Trip Override Issues 71152B In May 2019, a generator trip and subsequent reactor cutback occurred. The steam generators experienced a level deviation that placed the feedwater control system into manual. Westinghouse based the level deviation setpoint on engineering judgement to protect against gross failure in a level measurement and to place feedwater control into manual for further investigation. During this event the operators were expecting the feedwater control system to perform the reactor trip override which would have repositioned the feedwater regulating valves to predetermined positions and limit feedwater pump speed.

While the feedwater pumps were limited as expected the feedwater regulating valves did not reposition due to being in manual from the level deviation.

To correct the identified condition a modification was to be implemented to establish a level deviation setpoint that would not shift feedwater control to manual following reactor power cutbacks. The modification was implemented under engineering change 86323.

Condition report CR-WF3-2020-05623 documented that during the implementation of engineering change 86321, the prescribed level deviation setpoint of 15 percent could not be reached in the deviation module. Field change request 88313 evaluated a lower increase in level deviation setpoint to 10 percent due to hardware limitations that were not previously identified. The field change request concluded that a change to a lower set point did not bound the 2019 event.

The site failed to update the adverse condition analysis performed for the 2019 event with the information from field change request 88313 and this allowed the site to remain susceptible to this condition even if deviations of the same magnitude as the 2019 event occurred.

Condition report CR-WF3-2021-00483 documented that the margin gained implementing engineering change 86321 was less than originally evaluated in engineering changes 86321, 86322 and 86323. It was noted that the condition had been resolved during the outage when the engineering changes were implemented, and the intent of the CR-WF3-2021-00483 was to see if additional margin could be gained. This was a missed opportunity to correct the issue.

The condition identified as the physical cause of the 2019 event still existed at the site on March 21, 2024, when the reactor tripped due to a transformer fire again having the feedwater control to go to manual due to SG level deviation. The vulnerability identified in 2019 was not adequately corrected or mitigated.

Based on the recent event in 2024, the licensee is taking corrective actions to eliminate the level deviation setpoint to prevent the feedwater regulating valves from being placed into manual.

Observation: Deficient Actions to Correct Severity Level IV Violation for Not Providing Complete and Accurate Information to the NRC 71152B In January of 2024, the site received a Severity Level IV NCV 05000382/2023010-01 for the failure to verify the adequacy of design of the alternate source term dose consequence analysis and the failure to provide complete and accurate information to the NRC. The licensee documented this violation in CR-WF3-2023-18180. While the site has been performing corrective actions for the technical piece of the violation, it was apparent during the inspectors' review that no corrective actions were taken for the failure to provide complete and accurate information.

The inspectors noted during the review of the corrective actions performed to date that the licensee did not prioritize addressing the failure to provide complete and accurate information and focused the majority of their evaluation in resolving the technical issue. As a result, the licensee didn't have proper corrective actions to address that part of the violation.

The licensee has not completed all corrective actions associated with the condition report and recently added an additional corrective action (CA-07), due to NRC questions during the inspection, to CR-WF3-2023-18180 to document the actions that will be taken to address the failure to provide complete and accurate information to the NRC.

EXIT MEETINGS AND DEBRIEFS

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

On October 24, 2024, the inspectors presented the biennial problem identification and resolution inspection results to Joseph Sullivan, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

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v t e Inspection Report - Waterford - 2024010
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2024 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 10(0) 11(0) 12(0) 13(0) 40(0) Security 401(0) 402(0) 403(0) 420(0) Operator Exams Emergency Planning Other

IR 05000382/2024010

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SUBJECT:

Dear Joseph Sullivan:

Enclosure:

Inspection Report

SUMMARY

List of Findings and Violations

Additional Tracking Items

INSPECTION SCOPES

OTHER ACTIVITIES - BASELINE

71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)

INSPECTION RESULTS

Description:

Performance Assessment:

Enforcement:

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Enforcement:

Description:

Performance Assessment:

Enforcement:

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Performance Assessment:

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Enforcement:

EXIT MEETINGS AND DEBRIEFS

DOCUMENTS REVIEWED

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