IR 05000315/2023012
| ML23172A169 | |
| Person / Time | |
|---|---|
| Site: | Cook  (DPR–058, DPR–074) |
| Issue date: | 06/22/2023 |
| From: | Nestor Feliz-Adorno NRC/RGN-III/DORS/ERPB |
| To: | Lies Q Indiana Michigan Power Co |
| References | |
| IR 2023012 | |
| Download: ML23172A169 (1) | |
Text
SUBJECT:
DONALD C. COOK NUCLEAR PLANT - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000315/2023012 AND 05000316/2023012
Dear Q. Shane Lies:
On May 22, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Donald C. Cook Nuclear Plant 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 and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for problem identification and resolution programs.
Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.
The team also evaluated the stations processes for use of industry and NRC operating experience information and the effectiveness of the stations audits and self-assessments.
Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.
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.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
June 22, 2023 If you contest the violation or the significance or severity of the violation 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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at Donald C. Cook Nuclear Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at Donald C. Cook Nuclear Plant.
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, Néstor J. Féliz Adorno, Chief Engineering and Reactor Projects Branch Division of Operating Reactor Safety Docket Nos. 05000315 and 05000316 License Nos. DPR-58 and DPR-74
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000315 and 05000316
License Numbers:
Report Numbers:
05000315/2023012 and 05000316/2023012
Enterprise Identifier:
I-2023-012-0008
Licensee:
Indiana Michigan Power Company
Facility:
Donald C. Cook Nuclear Plant
Location:
Bridgman, MI
Inspection Dates:
May 01, 2023 to May 22, 2023
Inspectors:
M. Gangewere, Reactor Inspector
K. Kolaczyk, Reactor Operations Engineer
E. Magnuson, Reactor Inspector
E. Sanchez Santiago, Senior Project Engineer
Approved By:
Néstor J. Féliz Adorno, Chief
Engineering and Reactor Projects Branch
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 Donald C. Cook Nuclear Plant, 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 Establish Acceptance Criteria for Technical Specification Surveillance Requirement for Source Range Neutron Flux Monitoring Channel Check Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000315,05000316/2023012-01 Open/Closed
[P.2] -
Evaluation 71152B The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to include appropriate quantitative or qualitative acceptance criteria in procedure 2-OHL-4030-SOM-042 to demonstrate compliance with Technical Specification (TS) Surveillance Requirements (SR) 3.3.1.1 and 3.3.8.1. Specifically, the licensee failed to include acceptance criteria for determining the TS SR channel checks of source range neutron flux monitoring channels had been satisfactorily accomplished.
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)
- (1) The inspectors performed a biennial assessment of the effectiveness of the licensees Problem Identification and Resolution program, use of operating experience, self-assessments and audits, and safety-conscious work environment.
Problem Identification and Resolution Effectiveness: The inspectors assessed the effectiveness of the licensees Problem Identification and Resolution program in identifying, prioritizing, evaluating, and correcting problems. The inspectors also conducted a 5-year review of the Nuclear Instrumentation System and evaluated the corrective actions for the following non-cited violations, minor violations, and findings: NCV 2021011-01, "Incorrect Valve Design and Bearing Material Assumed for Safety Related Butterfly Valves 1-WMO-733/737 and 2-WMO-734/738," NCV 2022001-01, "Fire Zone Separation Not Maintained," and FIN 2022003-002, "Failure to Ensure Correct Operation of the Meteorological Tower."
Operating Experience: The inspectors assessed the effectiveness of the licensees processes for use of operating experience.
Self-Assessments and Audits: The inspectors assessed the effectiveness of the licensees identification and correction of problems identified through audits and self-assessments.
Safety-Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety-conscious work environment.
INSPECTION RESULTS
Assessment 71152B Based on the samples reviewed, the team concluded that the licensee's implementation of the Corrective Action Program was generally effective and supported nuclear safety.
Effectiveness of Problem Identification:
Based on the samples reviewed, the team concluded that the licensee continued to identify issues at a low threshold and appropriately entered these issues into the Corrective Action Program. The team determined that the licensee usually entered problems into the Corrective Action Program completely and accurately. The inspectors determined that the station has identified negative trends such as contractor oversight and engineering rigor. Though the site was generally effective at identifying negative trends that could potentially impact nuclear safety, the team identified multiple areas where trends were not formally documented, such as nuclear instrumentation failures, single point vulnerability issues, and impacts of steam leaks on the plant, including operations. In addition, the licensee used the Corrective Action Program to document instances in which previous corrective actions were ineffective or were inappropriately closed.
The team also noted that some deficiencies were identified by external organizations, including the NRC, that had not been previously identified by licensee staff and were subsequently entered into the Corrective Action Program. In addition, the licensee also utilized Corrective Action Program support processes to identify problems, including the self-assessment and audit process, and the Operating Experience Program. For example, the licensee performed department self-assessments and quality assurance audits to identify issues in station processes. Similarly, the licensee screened issues from both NRC and industry operating experience and entered them into the Corrective Action Program when they were applicable to the station.
The team performed a 5-year review of the nuclear instrumentation system. As part of this review, the team interviewed the system engineer, reviewed the plant health report, and reviewed selected corrective actions and condition evaluation documents. The team identified a non-cited violation for the failure to establish acceptance criteria for source range nuclear instrumentation surveillances. This violation is documented in this report.
Effectiveness of Prioritization and Evaluation of Issues:
Based on the samples reviewed, the team determined that licensee performance was generally effective at prioritizing and evaluating issues commensurate with the safety significance of the identified problem. The Initial Screening Committee and the Management Screening Committee meetings were generally thorough and intrusive in reviewing issues and prioritizing actions. In addition, the team observed a healthy dialogue between the members of these committees and the members challenged each other when dispositioning issues. However, the team identified multiple examples of issues that were classified as NCAP that should have been classified as CAP or CARC. This observation is documented in more detail in the Results section of this report.
In general, once a degraded or non-conforming condition was identified, the Corrective Action Program directed an equipment operability or functionality review to be performed. As a result, most of the samples reviewed were evaluated in a timely manner. However, the team noted multiple examples where an extent of condition review did not completely consider the impact of the identified condition on other similar equipment in the plant. This observation is documented in more detail in the Results section of this report.
Effectiveness of Corrective Actions:
Based on the samples reviewed, the team determined that the licensee was generally effective in corrective action implementation. In general, corrective actions for deficiencies that were safety significant were implemented in a timely manner. Problems identified using a root cause or other cause methodologies were resolved in accordance with Corrective Action Program requirements. The team determined that the licensee generally assigned corrective actions that were effective and timely for NRC identified issues and licensee event reports (LERs). However, the team identified an example where the licensee identified an error in a calculation used for determining gas accumulation acceptance criteria and did not establish short term corrective actions in accordance with procedures 12-EHP-5043-EDC-001, "Evaluation of Degraded/Nonconforming Conditions," and PMP-7030-CAP-002, "Condition Report Conduct and Closure." The details associated with this issue are documented in a minor performance deficiency in this report.
Assessment 71152B The team determined that the licensee's utilization of operating experience was generally effective. The licensee screened industry and NRC operating experience information for applicability to the station. When applicable, actions were developed and implemented to prevent similar issues from occurring. Operating experience lessons learned were communicated and incorporated into plant operations. The team did not identify any concerns in this area.
Assessment 71152B The team determined that the licensee's self-assessments and audits were generally effective. The licensee conducted department self-assessments and nuclear oversight audits periodically throughout the organization. These assessments and audits were generally effective in identifying issues and opportunities for improvement at an appropriate threshold.
The self-assessments and audits reviewed by the team identified issues that were not previously known, including issues within the Corrective Action Program itself. The licensee's Nuclear Oversight (NOS) also identified deficiencies in the licensee's processes, which were addressed through the Corrective Action Program. The team did not identify any concern in this area.
Assessment 71152B The team reviewed the results from the 2023 Employee Concerns Annual Assessment and multiple 2021 and 2022 departmental safety culture surveys. The team also conducted one-on-one interviews with 21 licensee staff concerning the effectiveness of the Corrective Action Program, the ability to raise issues, and the freedom from potential retaliation for raising issues. The team did not identify any impediment to the establishment of a safety-conscious work environment.
In general, the licensee's staff was aware of and familiar with the Corrective Action Program and other processes to raise nuclear safety concerns, such as the Employee Concerns Program. Licensee staff indicated they could raise nuclear safety concerns without fear of retaliation. The team did not identify examples of retaliation for raising nuclear safety concerns. The licensee staff interviewed believed that operational issues and issues with high safety significance were being appropriately addressed in a timely manner.
Failure to Establish Acceptance Criteria for Technical Specification Surveillance Requirement for Source Range Neutron Flux Monitoring Channel Check Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000315,05000316/2023012-01 Open/Closed
[P.2] -
Evaluation 71152B The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to include appropriate quantitative or qualitative acceptance criteria in procedure 2-OHL-4030-SOM-042 to demonstrate compliance with Technical Specification (TS) Surveillance Requirements (SR) 3.3.1.1 and 3.3.8.1. Specifically, the licensee failed to include acceptance criteria for determining the TS SR channel checks of source range neutron flux monitoring channels had been satisfactorily accomplished.
Description:
Donald C. Cook Nuclear Plant had a total of two source range neutron flux monitoring channels per Unit. TS 3.3.1, Reactor Trip System Instrumentation, required two source range neutron flux monitoring channels to be operable in mode 2 below the P-6 (intermediate range neutron flux) interlock, and modes 3, 4, 5 with rod control system capable of rod withdrawal or one or more rods not fully inserted. Technical Specification SR 3.3.1.1 required a channel check to be performed. Similarly, TS 3.3.8, Boron Dilution Monitoring Instrumentation (BDMI), required two source range neutron flux monitoring channels to be operable in modes 3, 4, and 5. TS SR 3.3.8.1 required a channel check to be performed.
The inspectors reviewed multiple action reports (ARs) for the Unit 2 source range neutron flux monitoring channels. Specifically, the inspectors reviewed AR 2021-4555, AR 2021-5609, AR 2022-7668, and noted deviations between the two source range channels were inconsistently evaluated. For example:
AR-2021-4555 documented 2-N-31 trended at a lower magnitude than 2-N-32. This issue was classified as Non-Corrective Action Program (NCAP) item. The actions taken included declaring 2-N-31 operable and completing an NCAP evaluation.
AR-2021-5609 documented 2-N-31 read a decade lower than 2-N-32. This issue was classified as a Condition Adverse to Quality (CAQ). The actions taken included declaring 2-N-32 inoperable and completing a maintenance rule evaluation.
AR 2022-7668 documented 2-N-31 read lower than 2-N-32. This issue was classified as NCAP. The actions taken included declaring 2-N-31 inoperable and completing a Failure Investigation Process. An assignment for an NCAP evaluation was performed to define a proper channel check between N-31 and N-32 and a lessons learned communication to licensed operators was initiated. However, no actions from either assignment were taken.
The inspectors also reviewed General Tracker 2021-6735. It stated, unlike most other indications available in the control room, there is no standard by which a source range is measured against. As a result, the licensee provided training to operators on nuclear instrumentation source range theory to assist in operability determination. The training was completed in January of 2022.
Technical Specification Bases Revision 60 for SR 3.3.1.1 and 3.3.8.1 stated, Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A Channel Check will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each channel calibration. The bases further stated, Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
The licensee established procedure 2-OHL-4030-SOM-042, Revision 56, Unit 2 Tours - U2 CR M3&4 Shift Chks, to implement TS SR 3.3.1.1 and 3.3.8.1 channel checks.
The inspectors noted that the procedure did not include quantitative or qualitative acceptance criteria for the channel checks of source range detectors 2-N-31 and 2-N-32. In contrast, the procedure included acceptance criteria for the intermediate range excore and power range excore detectors, including a minimum, maximum and/or maximum difference between channels values. Licensee procedure PMP-4030-EXE-001, Revision 28, Conduct of Surveillance Testing, defined acceptance criteria as parameters against which the collected data is to be compared to determine if: 1. Test completion is satisfactory. 2. Equipment OPERABILITY conditions are satisfied. 3. Technical Specifications are satisfied. 4. Design Basis criteria are adhered too.
The absence of source range detector acceptance criteria resulted in inconsistent operability determinations and follow up actions due to operator judgement variations from crew to crew.
The failure to include acceptance criteria to ensure TS SR 3.3.1.1 and 3.3.8.1 channel checks were completed satisfactorily was determined to also be applicable to the Unit 1 procedure 1-OHL-4030-SOM-029, Revision 59, Unit 1 Tours - U1 CR M3&4 Shift Chks, for the source range detectors.
Corrective Actions: The licensee entered this issue into their Corrective Action Program and planned to establish acceptance criteria in the affected TS SR procedures.
Corrective Action References: AR 2023-3814, "Undefined Acceptance Criteria for Source Range Channel Check"
Performance Assessment:
Performance Deficiency: The licensees failure to include quantitative or qualitative acceptance criteria for TS SR channel check implementing procedures of the source range neutron flux monitoring channels was contrary to 10 CFR 50 Appendix B, Criterion V, and 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 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 failure to establish acceptance criteria for the source range detectors TS SR channel checks did not ensure the detectors would continue to be able to perform their safety function.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because they answered No to the Reactor Protection System (RPS) question in exhibit 2, Mitigating Systems Screening Questions.
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 licensee did not thoroughly evaluate and resolve the absence of source range detector acceptance criteria when they identified there was a need to provide source range nuclear instrumentation calibration training to all operators to assist in operability determinations since there was no acceptance criteria for these instruments.
Enforcement:
Violation: Title 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings requires, in part, that instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Contrary to the above, as of May 15, 2023, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in procedures for determining that important activities have been satisfactorily accomplished. Specifically, the licensee failed to include acceptance criteria to determine the channel checks required by TS SR 3.3.1.1 and 3.3.8.1 were satisfactorily accomplished in procedures 2-OHL-4030-SOM-042 Revision 56, and 1-OHL-4030-SOM-029 Revision 59.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Minor Performance Deficiency 71152B Minor Performance Deficiency: During their review of open corrective actions the inspectors reviewed condition report AR-2018-7809, "Error in AEP-15-46." This AR documented that during a revision to the waterhammer analysis in report AEP-15-45, "Emergency Core Cooling System, Residual Heat Removal System and Containment Spray System Gas Accumulation Evaluation for D.C. Cook Units 1 and 2," it was identified that the vendor incorrectly modeled the discharge piping. This resulted in incorrect acceptance criteria for the waterhammer analysis for the gas accumulation program. The acceptance criteria at the time was 1.3 ft3. The vendor preliminary results expected the final acceptance criteria would be between 0.75 ft3 and 0.9 ft3.
Subsequently, the inspectors reviewed multiple ARs for voids identified in 2021 and 2022.
These ARs stated that the acceptance criteria was 1.3 ft3. The inspectors questioned whether the correct acceptance criteria was being used in the interim until the final corrective actions were taken to address the calculational error. Based on their review the inspectors identified the licensee continued to use 1.3 ft3 as the acceptance criteria with no documented justification to address the non-conforming condition. Procedure 12-EHP-5043-EDC-001, "Evaluation of Degraded /Nonconforming Conditions," Revision 31, Section 3.6.3 stated, "Determine Interim Actions for TS [Technical Specification] SSCs [structures, systems and components] and Non-TS SSCs Performing TS Support Functions." Procedure PMP-7030-CAP-002, "Condition Report Conduct and Closure," Revision 43, Section 3.5.3 stated, "Prior to approval, consider what impact the new due date could or would have on the list below: c. Interim Corrective Actions that may be required or are in place." Contrary to these self-imposed standards, the licensee did not determine interim actions for the identified non-confirming condition and in addition extended the due date for correcting the calculation multiple times without consideration of any interim corrective actions for the identified condition. The inspectors determined this was a performance deficiency and the licensee documented this issue in AR 2023-3844.
Screening: The inspectors determined the performance deficiency was minor. The inspectors determined the performance deficiency was of minor significance because in accordance with gas accumulation inspection procedures 1-EHP-4030-108-001A, "Train 'A' Monitoring and Trending of Gas Accumulation in ECCS," Revision 4, 1-EHP-4030-108-001B, "Train 'B' Monitoring and Trending of Gas Accumulation in ECCS," Revision 3, and 1-EHP-4030-108-004, "Outage Monitoring and Trending of Gas Accumulation in ECCS,"
Revision 16, the licensee would have maintained the identified voids below 20 percent of the acceptance criteria. This value is below the recommended interim acceptance criteria documented in AR 2018-7809.
Observation: Multiple instances of misclassification of condition reports 71152B The team reviewed corrective action documents and observed station review committee meetings, as well as management review committee meetings. The inspectors questioned the classification of the following condition reports as NCAP:
AR 2022-8824, "Missed verification steps on 2-IMO-220" - This AR stated the station sent a safety-related valve stem and nut to an offsite vendor. A licensee procedure required measurement of the components after machining and prior to installation.
This action was not performed and documented as a potential nonconformance.
AR 2023-3504, "1-PP-26N Oil analysis indicates a possible oil mix" - This AR documented that an oil analysis indicated a mix of the correct oil, and an unknown oil was used in the safety-related Unit 1 North safety injection pump.
AR 2023-3493, "CGID Evaluation Error Corrected" - This AR documented an error in a commercial grade dedication evaluation.
The justification for all of these ARs to be classified as NCAP was that there was no operability impact. The inspectors noted that procedure PMP-7030-MOP-001, "Corrective Action Program Management Oversight Process," defined a condition adverse to quality (CAQ) as a failure, malfunction, deficiency, deviation, effect or nonconformance associated with the performance of an activity affecting the safety-related function of a structure, system or component. This procedure also defined a condition adverse to regulatory compliance (CARC) as "A condition in which the licensee is not in conformance with NRC regulations; a failure to comply with a docketed commitment made to the NRC; a noncompliance with the licensees Quality Assurance Program that does not consequently affect the safety-related function of a structure, system or component." It also stated that "Conditions Adverse to Regulatory Compliance are addressed within the licensee Corrective Action Program." The inspectors noted using operability as the justification for classifying identified conditions was not in accordance with their established definitions. The importance of correctly classifying identified conditions is to ensure issues are fully identified, appropriate corrective actions assigned, trending of issues, among other actions specific to each classification option. The licensee reviewed the classification of the items questioned by the team and reclassified them. The inspectors did not identify any safety concerns associated with this observation.
Observation: Evaluation of Extent of Condition 71152B Prior to the inspection, the licensee identified multiple examples of concluding their review of identified issues without performing a thorough review. The identified examples resulted in repeat failures and issues. During the inspection, the team identified multiple examples of the licensee not fully characterizing extent of condition, including:
The licensee identified worn u-bolts on a 1/2 sampling line connected to the RCS Loop #1 hot leg during a routine walkdown of the Unit 1 containment. This discovery was documented in AR 2022-3764, AR 2022-3781, and AR 2022-3786. The station initially assigned a discrepant condition evaluation action (DCE) to fully evaluate the degraded condition of the sample line and u-bolts. Initial evaluation determined the cause of the loose u-bolts and fretting damage to the pipe to be vibration of the system. The DCE was cancelled when replacement of the u-bolts and affected piping was added into the outage work scope. The licensee did not perform an extent of condition on other similar sampling lines to determine if other lines had the same vibration induced degradation.
In 2020, pressurizer spray valve positioner 2-NRV-164-PU failed due to a previously unidentified single point vulnerability (SPV), causing the associated pressurizer spray valve, 2-NRV-164, to fail fully open. The resulting increase in pressurizer level led to a complicated scram with associated automatic safety injection. The licensee identified that the failed position, and others of the same type, were inappropriately removed from the SPV program in 2018. The station did not perform a review of their SPV program changes to determine if additional components had been inappropriately removed at that time. Subsequently, in 2021, the 2-XJ-113-5 bellow from the moisture separator reheater failed during operation, which led to operators inserting a reactor scram. The licensees causal evaluation (AR 2021-5596) determined that the failed bellows should have been classified as a single point vulnerable component. One of the initial corrective actions assigned as part of the root cause evaluation was to perform a review of all single point vulnerability related changes and evaluations over the previous 10 years. This assignment was removed by the corrective action review board and replaced with an ongoing generic review of risk significant engineering changes. In both cases, when latent single point vulnerabilities caused plant transients, the station did not evaluate their single point vulnerability program for further issues.
The inspectors did not identify any safety concerns associated with this observation.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On May 22, 2023, the inspectors presented the biennial problem identification and resolution inspection results to Q. Shane Lies and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Calibration
Records
AR 2022-9291
Unit 2 Reactor Trip
11/10/2022
AR 2022-2860
ACE for Missed LHRA Posting During Sluice Activities
04/07/2022
AR 2010-11114
Exposed Insulation Found During U2C19 Recirc Sump
Walkdown
10/19/2010
AR 2015-15397
CDBI Identified Violations
11/30/2015
AR 2018-6573
Roof Leak
06/24/2018
AR 2019-11423
2-NRI-32 is Not Responding Correctly and is Inoperable
11/14/2019
AR 2020-0312-9
MRE: N-44 Failed to 102% in U2 Caused a Unexpected
CR Alarm
01/15/2020
AR 2020-1423
MRule (a)(1) Process for Unit 2 Nuclear Instrumentation
2/12/2020
AR 2020-3739
2-NRI-32 is Slowly Failing Low
05/03/2020
AR 2020-6997
Unit 2 Source Range Instrumentation
09/07/2020
AR 2020-8407
Filters do Not Literally Meet Implied LRA Requirements
10/09/2022
AR 2020-8661
Update the SPV List and FMEA's for Charging & Spray
Valves
10/15/2020
AR 2021-0040
Perform Common Cause Evaluation of Pressurizer PORV
Leakage
01/04/2021
AR 2021-0040
Perform Common Cause Evaluation of Pressurizer PORV
Leakage
01/02/2021
AR 2021-10263
Clearance Request Submitted by Maintenance Planners
2/16/2021
AR 2021-10275
1-CCW-256 Removed from U1C31 Scope - Perform Risk
Eval
2/16/2021
AR 2021-2838
Vulnerabilities Introduced During Design Modifications.
04/05/2021
AR 2021-3483
2-NRI-32 Not Reading Properly
04/24/2021
AR 2021-4555
2-NRI-31 and 2-NRI-32 Not Trending at the Same
Magnitude.
05/17/2021
AR 2021-5609
U2 Source Range Detector 2-NRI-21 is Inoperable
06/23/2021
AR 2021-5690
Unreliable Performance of Unit 2 Source Range Detectors
06/24/2021
AR 2021-5745
2-N-32 is Failing Low.
06/27/2021
Corrective Action
Documents
AR 2021-6327
CR's Not Written for Evaluations on Failed Safety Related
Parts
07/21/2021
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
AR 2021-7098
1-WMO-737 Triple Offset Instead of Double Offset
08/17/2021
AR 2021-7594
1-WMO-733 As-Left Un-Seating Torque Above VDS Criteria
09/07/2021
AR 2021-8277
NRC POV Violation Internal Valve Maintenance
10/04/2021
AR 2021-8808
Discrepancy Identified in SFP Inventory Detailed Description
10/26/2021
AR 2022-0168
EACE for Unit 1 Aux Cable Vault CO2 Actuation
01/06/2022
AR 2022-0168
Unit 1 Aux Cable Vault CO2 Actuated
01/06/2022
AR 2022-0195
1-HV-ACE-S1-FD Failed to Close.
01/06/2022
AR 2022-0198
Inadequate Guidance Regarding TRM Actions
01/07/2022
AR 2022-0936
EDG 2R Lifter Damaged
2/01/2022
AR 2022-10001
Evaluate Unit 2 Pressurizer Heaters and Spray Bypass Valve
2/11/2022
AR 2022-1042
Freezing Frazzle Ice Conditions
2/23/2022
AR 2022-1272
RVLIS Uncertainty Potentially Non-Conservative
2/15/2022
AR 2022-2860
Demin Room Not Posted as LHRA During Resin Sluice as
LHRA During Resin Sluice
04/07/2022
AR 2022-3764
1-RC-101-L1 Line Support Wearing Out
04/29/2022
AR 2022-3786
Second Worn U-bolt Downstream of 1-RC-101-L1
04/29/2022
AR 2022-3870
CRDM Cables Dangling out of Window into the Reactor
Cavity
05/02/2022
AR 2022-4033
Weld Failed Dye Penetrant Exam
05/07/2022
AR 2022-41
Corrosion in Fire Protection Piping
03/27/2013
AR 2022-4255
Gas Void at 1-CS-353
05/14/2022
AR 2022-4994
U2 T/C #43 (Core Location F05) Erratic Readings
06/08/2022
AR 2022-5785
Removal of CST and Connected Piping from ISI Program
07/13/2022
AR 2022-6282
Failure to Implement/Maintain Procedures for the MET Tower
08/04/2022
AR 2022-6828
Unit 1 Reactor Trip - RCP 13 Trip
08/28/2022
AR 2022-6921
SCD-PE-EVALUATION-100001022 Came to Incorrect
Conclusion
08/30/2022
AR 2022-7538
Track A(1) Process for RCS-09 IAW 12-EHP-5035-MRP-001
09/27/2022
AR 2022-7668
2-N-31 Failed Low
10/01/2022
AR 2022-7740
Critical Parameters Found out of Tolerance
10/02/2022
AR 2022-7824
Unacceptable Conditions Identified During QC Inspection.
10/04/2022
AR 2022-7827
Tube Cleaning Brushes Found in 2-HE-15E
10/04/2022
AR 2022-7862
CR Not Initiated Same Shift as Condition
10/05/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
AR 2022-7876
Lifting and Rigging Knowledge Gap
10/05/2022
AR 2022-7877
Radiation Worker Practices
10/05/2022
AR 2022-7924
Shield Wall Plate Bolt Pulling from Wall
10/06/2022
AR 2022-7925
DRT217 VA Door Found Unsecure
10/06/2022
AR 2022-7943
Lost Pen with Lanyard in Unit 2 Upper Ice Condenser
10/07/2022
AR 2022-7967
FME Earplug End Broken Off in Ice Condenser - Retrieved
10/07/2022
AR 2022-7975
BHI Radiation Worker Practices
10/07/2022
AR 2022-8029
Lube Oil Found without CCM Permit
10/09/2022
AR 2022-8033
Dropped Object in Upper Ice Condenser
10/09/2022
AR 2022-8046
Instrument Tubing Pulled Out at 2-HE-47-CDN
10/09/2022
AR 2022-8088
10/10/2022
AR 2022-8107
Dropped Object During Containment Flange Removal.
10/10/2022
AR 2022-8158
Tube Cleaning Brush Lost in Drain to TRS
10/12/2022
AR 2022-8160
Working Within 6' of Leading Edge w/o Fall Protection
10/12/2022
AR 2022-8164
Maintenance Permit Procedure Performance Errors
10/12/2022
AR 2022-8206
Master Lee Lifting and Rigging
10/13/2022
AR 2022-8215
Unit 2 MSR TK-97 Support Baseplate Anchor Bolt Broken
Off
10/13/2022
AR 2022-8267
2-ICR-6 Accumulator Tank Sample TR B CIV Failed LLRT
10/15/2022
AR 2022-8278
Peer Inspector without QC-Q-0001
10/15/2022
AR 2022-8300
Work Group Not Complying with PMP-2270-CCM-001
10/16/2022
AR 2022-8380
Certified Instruments Issued for More Than 10 Days.
10/18/2022
AR 2022-8390
Temporary Power Cords Not Routed
IAW PMP-2281-SWP-001
10/18/2022
AR 2022-8473
Ice Machine Hose Leaks
10/20/2022
AR 2022-8509
Unacceptable Weld
10/20/2022
AR 2022-8545
Partial Discharge Activity Observed on RCP23's Cables.
10/21/2022
AR 2022-8728
Action Not Reported During Shift
10/26/2022
AR 2022-8772
Field Work Performed Prior to Design Approval
10/25/2022
AR 2022-8805
Combustibles Left Unattended in Welding Work Area
10/28/2022
AR 2022-8824
Missed Verification Steps on 2-IMO-220
10/29/2022
AR 2022-9180
2-DRV-407 Leaks By
11/07/2022
AR 2022-9190
Unit 2 N-32 Reading 5 Cps when De-Energized.
11/08/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
AR 2022-9291
Unit 2 Reactor Trip
11/10/2022
AR 2022-9308
Unit 2 Plant Heatup Issues
11/11/2022
AR 2022-93921
ODE Has Been Requested for U1 TDAFWP Room
Temperature
11/15/2022
AR 2022-9698
Unit 1 RCS In-Leakage Trend
11/29/2022
AR 2023-0504
RCE for Unit 2 Reactor Trip Failed CARB
01/18/2023
AR 2023-0812
1-WFI-743 Needs to be Replaced
01/26/2023
AR 2023-0854
2-CF-128 Check Valve is Leaking By
01/27/2023
AR 2023-1862
2-WMO-703 Failed to Auto Open During Pump Start.
03/04/2023
AR 2023-2457
Unit 1 Has Entered Tier 1 Actions for RCS Leakage
03/14/2023
AR 2023-3294
EC-58174 for 2-MRV-213-F Issues
04/26/2023
GT 2019-4420
Unusually Low Reading on Source Range Nuc.
Instrumentation
04/26/2019
GT 2020-3839
Knowledge Gap Identified in Calibrating Source Range NI's
05/06/2020
GT 2021-0078-
2,3,4,5,6
ESYE General Tracker - 2021
01/04/2021
GT 2021-1619
Quick Hit Self Assessment on NFPA 805 Fire Protection
Program
2/18/2021
GT 2021-2361
Update Isometric Drawing for 1-SV-78-AB1
03/16/2021
GT 2021-4250
NRC Information Notice 2021-01 DBAI PORV Inspections
05/11/2021
GT 2021-4925
Rx Trip Caused by Spurious Neutron Flux Signal
05/27/2021
GT 2021-5546
Westinghouse Nuclear Instrumentation Bypass Panel
06/21/2021
GT 2021-7069
NRC Information Notice IN 2021-03, OE Re: Derecho Event
08/16/2021
GT 2021-7529
Source Range High Flux Reactor Trip Bypassed When
Required
09/02/2021
GT 2021-9551
Industry Level 1 Clearance Events
11/23/2021
GT 2022-0636
QH for Repeat Maintenance
01/24/2022
GT 2022-10273
NSRB November 2022 Meeting Action Items
2/21/2022
GT 2022-1166
Manual Reactor Trip Due to an Electro-Hydraulic Oil Leak
2/10/2022
GT 2022-1694
STRIDE STI-22-01 RVLIS Surveillance Frequency Extension
03/02/2022
GT 2022-4161
Enertech Valve Design Basis Verification Qualification
05/11/2022
GT 2022-5363
Automatic Reactor Trip Following Loss of All Four RCPs
06/23/2022
GT 2022-5789
2-PP-145E East Diesel Fire Pump Inboard Pump Packing
07/13/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Temps
GT 2022-7741
Source Range NI Surveillance Tracking
10/02/2022
GT 2022-9150
NRC Regulatory Issue Summary 2022-02, Operational
Leakage
11/07/2022
0058180
Revise Drawing 1-CA-539-L1-3 per As-Built
Info\GT 2021-2361
Implementation of the Inservice Inspection Program
Boundaries Bases Document
Engineering
Changes
Unit 2 Control Air Filter Installation for
2-MRV-213/223/233/243
ARM-21-04-02
Equipment Trending Challenges
04/08/2021
I&C LRTP
I&C Long Range Training Plan Specialty/Task Refresher
22
ISA-S7.3-1975(R
1981)
Quality Standard for Instrument Air
NI Q1 2023
Nuclear Instrumentation System Walkdown Report
03/28/2023
NI Q2 2022
Nuclear Instrumentation System Walkdown Report
04/11/2022
NI Q3 2022
Nuclear Instrumentation System Walkdown Report
07/11/2022
NI Q4 2022
Nuclear Instrumentation System Walkdown Report
2/08/2022
Performance
Monitoring Plan
Nuclear Instrumentation System
2/14/2019
PRF230019
Enertech Valve Design Basis Verification Qualification
Q4-2020
Unit 2 Nuclear Instrumentation System Health Report
03/14/2020
Q4-2021
Unit 2 Nuclear Instrumentation System Health Report
03/21/2022
Q4-2021
Unit 1 Nuclear Instrumentation System Health Report
03/21/2022
Miscellaneous
U2 NI-01 (a)(1)
Action Plan Revision
11/10/2021
1-EHP-6040-164-
001
Unit 1 Control Air Sampling Test
1-IHP-4030-113-
29
Nuclear Instrumentation Source Range Channel Operational
Test and Calibration
1-IHP-4030-113-
29
Nuclear Instrumentation Source Range Channel Operational
Test and Calibration
Procedures
2-EHP-5036-
EQR-002
System Engineering
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-IHP-5040-
EMP-004
Plant Winterization and De-Winterization
2-EHP-6040-264-
001
Unit 2 Control Air Sampling Test
2-IHP-4030-213-
29
Nuclear Instrumentation Source Range Channel Operational
Test and Calibration
DTG-VSR-001
Design Vertical Slice Review Process
DTG-VSR-001
Design Vertical Slice Review Process
EHI-5054-ICT
Non-EQ Instrumentation Circuits Test Review Program
EHI-5054-SWD-
001
System Walkdowns
MHI-5024
Repeat Maintenance and Trending Process
PMI-5055
Winterization-Summarization
PMI-7030
Corrective Action Program
PMP-2291-PMT-
001
Work Management Post Maintenance Testing Matrices
PMP-4030-EXE-
001
Conduct Of Surveillance Testing
PMP-5055-001-
001
Winterization-Summarization
PMP-7030-CAP-
001
Action Initiation
PMP-7030-CAP-
2
Condition Report Conduct and Closure
PMP-7030-MOP-
001
Corrective Action Program Management Oversight Process
PMP-7030-MOP-
001
Corrective Action Program Management Oversight Process
PMP-7030-OPR-
001
PMP-7030-TND-
001
Trend Analysis
PMP-7032-FIP-
001
Failure Investigation Process
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
GT 2018-0522
Instrument Air Quality Quick-Hit Self-Assessment (QHSA)
GT 2019-1551
Quick Hit Self Assessment per PMP-7034-SAP-001 on Fluid
Leak Management
2/18/2019
GT 2021-9788-1
Assessment of Important Equipment Failures System
Engineering Department
2/08/2021
GT 2021-9791
Cable Aging and Monitoring QHSA
2/02/2021
GT-2015-13155
Create Action Plan to Improve Filtration for Analog
Positioners
10/09/2015
GT-2021-7403
Quick Hit Self Assessment per PMP-7034-SAP-001
Winterization Program
08/30/2021
Self-Assessments
NOS-21-08
Engineering, Design Control, In-Service
Inspections/In-Service Tests
09/15/2021
55342726
Replace Aged Channel I Triaxial Cables
2/23/2018
55543105-01
Non-EQ Instrumentation Circuits Cables Test Review
2/08/2023
55545156
WR for 2-NRI-32 Cable Inspection/Repair
10/04/2022
55562843
1-WMO-737 Valve Replacement per ACE AR 2021-1897
04/20/2022
55563578
Anomalies Identified by AMS During CHAR Testing
07/06/2021
5556461801
MTI, 2-IHP-4030-213-229, COT and Calibration CH1, N-31.
10/02/2022
5556461901
MTI-2-IHP-4030-213-229: COT and Calibration CH2 N-32
10/01/2022
C10022138001
22-3764, 1-RC-101-L1, Repair Tubing Section between
NSX-I01 & NRV-101
05/04/2022
Work Orders
C10046877
2-CF-128 Check Valve is Leaking By
2/28/2023