IR 05000333/2022004
| ML23076A122 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/20/2023 |
| From: | Erin Carfang NRC/RGN-I/DORS |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2022004 | |
| Download: ML23076A122 (1) | |
Text
March 20, 2023
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - INTEGRATED INSPECTION REPORT 05000333/2022004
Dear David Rhoades:
On December 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at James A. FitzPatrick Nuclear Power Plant. On February 1, 2023, the NRC inspectors discussed the results of this inspection with Garrick Olson, Plant Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.
Three 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.
A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as an NCV consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at James A. FitzPatrick Nuclear Power 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 I; and the NRC Resident Inspector at James A. FitzPatrick Nuclear Power 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 (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety
Docket No. 05000333 License No. DPR-59
Enclosure:
As stated
Inspection Report
Docket Number:
05000333
License Number:
Report Number:
Enterprise Identifier: I-2022-004-0032
Licensee:
Constellation Energy Generation, LLC
Facility:
James A. FitzPatrick Nuclear Power Plant
Location:
Oswego, NY
Inspection Dates:
September 30, 2022 to December 31, 2022
Inspectors:
E. Miller, Senior Resident Inspector
J. England, Resident Inspector
F. Arner, Senior Reactor Analyst
C. Bickett, Senior Reactor Analyst
S. Haney, Senior Project Engineer
M. Hardgrove, Senior Project Engineer
J. Kulp, Senior Reactor Inspector
J. Schussler, Senior Resident Inspector
Approved By:
Erin E. Carfang, Chief
Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at James A. FitzPatrick Nuclear Power 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. A licensee-identified non-cited violation (NCV) is documented in report section 71111.1
List of Findings and Violations
Failure to Perform Preventive Maintenance on Raw Water Manual Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000333/2022004-01 Open/Closed None (NPP)71111.12 The inspectors identified a Green finding and associated non-cited violation (NCV) of Technical Specifications (TS) 5.4.1(a), Procedures, for failure to establish preventive maintenance in accordance with station procedures for a safety-related component.
Specifically, emergency service water manual valve 46ESW-12A was not receiving preventive maintenance. As result, a stem to disc separation occurred causing a loss of emergency service water supply to the A train emergency core cooling systems.
Emergency Core Cooling Pump Switch Failures Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000333/2022004-02 Open/Closed
[P.2] -
Evaluation 71111.13 The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify a condition adverse to quality associated with main control room emergency core cooling system (ECCS)pump switches. Specifically, Constellation failed to identify a potential common failure mechanism following three failures of ECCS main control room pump switches. As a result,
Constellation did not properly classify significance following the third failure and although action items were established, a corrective action was not assigned in accordance with Constellation corrective action program requirements.
Lead Shielding Not Analyzed in Accordance with Station Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000333/2022004-03 Open/Closed None (NPP)71111.15 The inspectors identified a Green finding and associated non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for Constellation's failure to evaluate shielding in accordance with station procedures. Specifically, the shielding was in contact with reactor building closed loop cooling piping which is seismic class I and serves as part of the containment boundary.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000333/2022-001-00 Exhaust Drain Pot Line Filled with Water up to HPCI Turbine Due to Relay Failure 71153 Closed
PLANT STATUS
FitzPatrick began the inspection period offline for refuel and maintenance outage J1R25. On October 16, 2022, operators commenced a reactor startup. On October 19, 2022, FitzPatrick reached rated thermal power. After reaching rated thermal power, on October 19, 2022 operators reduced reactor power to 80 percent to perform a control rod pattern adjustment and restored reactor power to rated thermal power on October 20, 2022. On October 20, 2022, operators reduced reactor power to 80 percent to perform a control rod pattern adjustment.
Operators completed restoration to rated thermal power on October 21, 2022. On November 18, 2022, operators reduced reactor power to 65 percent to perform a control rod pattern adjustment. Power was restored to rated thermal power on November 19, 2022.
FitzPatrick remained at or near rated thermal power for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk-significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures for the following system:
- Emergency diesel generator rooms on December 20, 2022
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Decay heat removal system during refueling outage J1R25 shutdown conditions on October 3, 2022
Complete Walkdown (IP Section 03.02) (1 Sample)
- (1) 'A' containment atmosphere dilution system during maintenance on the 'B' system during the week of December 12, 2022
71111.05 - Fire Protection
Fire Area Walkdown and Inspection (IP Section 03.01) (2 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Motor generator set room, elevation 300', fire area/zone 1A/MG-1; and relay room, elevation 286', fire area/zone VII/RR-1, on December 1, 2022
- (2) Main control room and control room heating, ventilation and air conditioning equipment rooms, elevation 300', fire area/zone VII/CR-1, on December 29, 2022
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 03.01) (2 Samples)
The inspectors evaluated internal flooding mitigation protections in the:
- (1) Manholes 4A, 4B, and 5B on December 2, 2022
- (2) East and west crescent on December 27, 2022
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities - Nondestructive Examination and Welding Activities
(IP Section 03.01)
- (1) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from September 27, 2022 to October 3, 2022:
- Ultrasonic Examination of the N-3B Nozzle to Shell Weld (1R25-ISI-UT-023)
- Ultrasonic Examination of the N-3B-IR Nozzle Inner Radius UT (1R25-ISI-UT-039)
- Phased Array Ultrasonic Examination of N2KSE Nozzle to Safe End Weld ([[::JAF-3002-N2KSE|JAF-3002-N2KSE]])
- General Visual Examination of Torus Interior RB 227 to 272 (Torus Shell Above Water) (1R25-IWE-VT-002)
- General Visual Examination of Containment Shell and Process Piping Penetrations (Main Steam and Main Feed) in Steam Tunnel (RB 256-268)
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed the reactor startup from refueling outage J1R25 on October 16, 2022.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed Just-in-Time training for reactor startup and heatup on October 14, 2022.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Internal torus structure on October 3, 2022 (2)12ESW-12A, Manual Supply Isolation Valve to the West Crescent Unit Coolers, on December 12, 2022
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Elevated risk during control rod drive mechanism replacement on October 1, 2022
- (2) Elevated risk during reactor cavity draindown on October 10, 2022
- (3) Elevated risk during planned high-pressure coolant injection (HPCI) system maintenance during week of November 16, 2022
- (4) Elevated risk during blocked 'A' emergency service water flow to the west crescent unit coolers and 'A' low pressure coolant injection battery failure on December 3, 2022
- (5) Emergent risk during failure of 'B' core spray pump to start due to main control room switch issue on December 9, 2022
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)29AOV-86A, 'A' outboard main steam isolation valve following failure to close on October 3, 2022 (2)10MOV-25A, low pressure coolant injection inboard valve due to valve hammering issue on October 3, 2022
- (3) Lead shielding in contact with reactor building closed loop cooling piping on October 6, 2022
- (4) Control rod position indication due to non-functional lights on November 15, 2022
- (5) Control rod 14-39 overtravel verification following annunciator modification on November 18, 2022
- (6) 'A' emergency service water pump following failure of comprehensive pump test on December 2, 2022
- (7) 'B' emergency service water through-wall leak on December 28, 2022
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Temporary Modification: Seal Weld Design Evaluation - Reactor Core Isolation Cooling System Valve 13MOV-21 on October 5, 2022
- (2) Temporary Modification: Control Rod 14-39 Overtravel Annunciator Defeated on November 16, 2022
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (11 Samples)
The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:
(1)10MOV-25A, Low Pressure Coolant Injection Isolation Valve Test following over-thrust event, on September 28, 2022 (2)23MOV-14, High-Pressure Coolant Steam Supply Isolation Valve Test following wedge and seat repair on October 3, 2022 (3)13MOV-21, Reactor Core Isolation Cooling Discharge Isolation Valve Test following inspection and seal weld on October 5, 2022 (4)29AOV-86A, 'A' Outboard Main Steam Isolation Valve Test following air pack replacement on October 5, 2022
- (5) Loss of Power and Loss of Coolant Accident Emergency Power Load Sequence Test following 'A' and 'C' emergency diesel generator digital reference unit replacement, on October 7, 2022
- (6) ST-6M, Standby Liquid Control Recirculation, Injection Test following squib valve replacement on October 8 and 9, 2022
- (7) 'G' safety relief valve control room switch replacement on October 9, 2022
- (8) ST-39H, RPV System Leakage Test (Hydro) following control rod drive replacements, on October 12, 2022
- (9) 'A' reactor feedwater pump high water level trip due to 31MOV-100A logic malfunction on October 16, 2022
- (10) HPCI system following preventive maintenance on November 18, 2022
- (11) 'B' core spray main control room switch following replacement on December 15, 2022
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated refueling outage J1R25 activities from October 1 to 17, 2022.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (3 Samples)
- (2) ESP-50.002, Floor Drain Flow Test - Plant Shutdown, on October 6, 2022
- (3) ST-2AJ, Residual Heat Removal Loop 'A' Containment Spray Headers and Nozzles Air Test (ISI), on October 5, 2022
FLEX Testing (IP Section 03.02) (1 Sample)
71114.06 - Drill Evaluation
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
- (1) The inspectors evaluated a simulator exercise that involved a loss of high-pressure injection, a loss of coolant accident, and an emergency depressurization on November 15,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (IP Section 02.04)===
- (1) October 1, 2021 through September 30, 2022
MS06: Emergency AC Power Systems (IP Section 02.05) (1 Sample)
- (1) October 1, 2021 through September 30, 2022
MS07: High-Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) October 1, 2021 through September 30, 2022
MS08: Heat Removal Systems (IP Section 02.07) (1 Sample)
- (1) October 1, 2021 through September 30, 2022
MS09: Residual Heat Removal Systems (IP Section 02.08) (1 Sample)
- (1) October 1, 2021 through September 30, 2022
MS10: Cooling Water Support Systems (IP Section 02.09) (1 Sample)
- (1) October 1, 2021 through September 30, 2022
===71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03)===
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) IR 04450965 - Central Programs Office (CPO) Oversight Actions, J1R25 Deferral of N-2F-SE-UT Exam
- (2) IR 04496616 - HPCI Turbine Water Accumulation Due to Relay Failure on October 31, 2022
- (3) IR 04443511 - Review of NRC Identified Trend of Failure to Adhere to Corrective Action Program Requirements to Document, Track, and Implement Actions
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the FitzPatrick corrective action program for trends that might be indicative of a more significant safety issue.
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000333/2022-001-00, Exhaust Drain Pot Line Filled with Water Up to HPCI Turbine Due to Relay Failure (ADAMS Accession No. ML22179A366). This LER was submitted and subsequently retracted by licensee correspondence titled "Retraction of LER: 2022-001, Exhaust Drain Pot Line Filled with Water up to HPCI Turbine Due to Relay Failure," dated July 15, 2022 (ADAMS Accession No. ML22196A392). The inspection conclusions associated with this LER are documented in this report under Inspection Results.
Personnel Performance (IP Section 03.03) (3 Samples)
- (1) The inspectors evaluated the loss of 115-kilovolt offsite power source Line 4, and the licensee's performance, on November 3, 2022.
- (2) The inspectors evaluated the HPCI alarm response for valve or motor overload or loss of control power, and the licensee's performance, on November 8, 2022.
- (3) The inspectors evaluated an inadvertent initiation of water curtain spray 5, and the licensee's performance, on November 23,
INSPECTION RESULTS
Very Low Safety Significance Issue Resolution Process: Failure to Perform General Visual Examination of Accessible Surfaces of Containment Penetrations Required by Paragraph 10 CFR 50.55a(g)(4)71111.08G This issue involves containment weld in-service examination requirements that warrant clarification via generic agency processes outside of inspection and assessment to address broader safety and regulatory concerns. The Very Low Safety Significance Issue Resolution was revised in August 2022 to provide for this option.
Description:
On September 30, 2022, the inspectors observed Constellations general visual examination of Class MC surface areas of flued head components at containment penetrations X-7A, X-7B, X-7C, X-7D, X-9A, X-9B, and X-8. The work order directed the examiners to perform a general visual examination of American Society of Mechanical Engineers (ASME) Category E-A, class E1.11 components (Accessible Surface Areas of the Containment Vessel Pressure Retaining Boundary) in the steam tunnel area of the reactor building, which is where the penetrations listed above are located. The inspectors noted that Constellation staff did not remove the blanket-style thermal insulation covering the penetrations and associated flued head components to allow access for the examiners to perform the general visual examination. Constellation examiners could only access an approximately 3-inch-wide area of each penetration that was not covered by blanket-style thermal insulation. The uninsulated area of the penetration did not contain the flued head circumferential weld. Constellation staff documented the examination of all accessible penetrations in Nondestructive Examination (NDE) Report 1R25-IWE-VT-016 with no recordable indications identified.
The inspectors inquired why the blanket-style thermal insulation was not removed prior to the examinations. The inspectors observed that blanket-style thermal insulation is draped over a component and is held in place by common commodity tie wire. Blanket-style insulation is commonly removed from components to provide access for examiners to perform examinations. Specifically, the inspectors questioned whether Constellation could credit the IWE general visual examination of the subject containment penetrations covered by insulation. In response to the inspectors questions, Constellation staff provided an email response documenting their position with regards to the containment accessible surface areas. The email stated, in part, It is the position of Constellation that containment surface areas covered by thermal insulation are not considered accessible for general visual examinations in accordance with Table IWE-2500-1, Examination Category E-A. The inspectors further found that the licensee staff could not retrieve prior inservice examination records of the flued head welds.
The inspectors reviewed pertinent drawings and observed Constellation's classification of these penetrations as inaccessible was different from their containment inservice inspection (CISI) boundary drawings that were in effect at that time. Specifically, drawing ISI-IWE-004, Sheet 1, designated the process piping penetrations and flued head component surface areas of these containment penetrations with a note stating, Accessible surfaces within IWE boundary, requiring inspection. The inspectors observed that Constellation's position, according to their email response to the inspector, was to not remove thermal insulation when performing General Visual examinations on containment accessible surfaces areas in accordance with Table IWE-2500-1, Examination Category E-A, as supported by ASME Code Section XI Interpretation XI-1-13-25. Constellation supported their position by stating, in part:
The requirements of Subarticle IWE-2310 apply when performing general visual examinations under Examination Category E-A, Item E1.11. Paragraph IWE-2310(c) states that 'visual examinations shall be performed, either directly or remotely, by line of sight from floors, platforms, walkways, ladders, or other permanent vantage points, unless temporary access is required by the inspection plan.' Removal of thermal insulation to perform the Examination Category E-A, Item E1.11 general visual examinations is not required because removal of the insulation would be a temporary measure to permit examination of surfaces that would otherwise be inaccessible for examination in accordance with the requirement of IWE-2310(c). Temporary access requiring removal of insulation is not specified in the ISI Plan. ASME Interpretation XI-1-13-25 [] confirms the intent of the code with respect to accessibility of containment surfaces covered by thermal insulation. [] This interpretation also supports Constellations conclusion that the performed examinations comply with ASME Code Section XI Subsection IWE. As such, an alternative to the requirements of Subsection IWE in accordance with 10 CFR 50.55a(z) is not needed.
The inspectors observed that Constellations position of considering the containment surface area covered by removable thermal insulation as inaccessible was based on ASME Code Interpretation XI-1-13-25, which stipulated it was not a requirement of IWE-1230 that containment surface covered by thermal insulation be considered accessible for general visual examination in accordance with Table IWE-2500-1 (E-A). However, the inspectors noted the Code inquiry (dated November 13, 2013) that prompted Interpretation XI-1-13-25 stated its purpose was to clarify whether removal of containment shell or liner plate thermal insulation panels is required in order to make containment surfaces accessible for visual examination in accordance with IWE-2500, Table IWE-2500-1, Examination Category E-A. The Code inquiry further explained, In many cases, this insulation material is not intended to be removed and is physically attached to the containment surface. Therefore, the inspectors, in consultation with staff of the Office of Nuclear Reactor Regulations (NRR),noted that Interpretation XI-1-13-25 was intended to address the accessibility for visual examination of containment shell or liner plate with insulation panels physically attached and not intended to be removed. However, Constellations general visual examinations of the containment penetrations and flued head weld were not on the containment shell or liner plate with insulation panels physically attached and not intended to be removed.
Procedure ER-JF-330-1001, James A FitzPatrick Nuclear Power Plant Station Inservice Inspection (ISI) Program Plan," Revision 2, delineated the code of record for the third 120-month CISI interval as the ASME Section XI Code, 2007 Edition through the 2008 Addenda. This program document was applicable for the third CISI interval, which was effective from August 1, 2017, through June 15, 2027. The current interval for the CISI was subject to the requirements of the 2007 Edition through the 2008 Addenda of the ASME Section XI, Subsection IWA, General Requirements, and Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants.
Subsection IWE, Subarticle IWE-1230, Accessibility for Examination, stipulated, in part, that the openings and penetrations of Class MC containment vessels, parts, and appurtenances remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant. This subarticle also stipulated, in part, that surface areas of Class MC containment vessels, parts, and appurtenances were considered inaccessible if visual access by line of sight from permanent vantage points was obstructed by permanent plant structures, equipment, or components.
The flued head components at containment penetrations X-7A, X-7B, X-7C, X-7D, X-9A, X-9B, and X-8, were classified by the licensee as Class MC and were subject to general visual examination requirements in accordance with Subsection IWE, Table IWE-2500-1, Examination Category E-A, Containment Surfaces. Table IWE-2500-1 (E-A), Item E1.11, Accessible Surface Areas, required general visual examination of all accessible surface areas in accordance with IWE-2310, Visual Examinations. Subarticle IWE-2310 as well as IWE-2311, General Visual Examinations, required general visual examinations to be performed in accordance with IWE-2500 and Table IWE-2500-1 (E-A) to determine the general condition of containment surfaces and detect evidence of degradation.
NRC Inspection Manual, Part 9900: Technical Guidance titled, American Society of Mechanical Engineers Boilers and Pressure Vessel Code, Sections III & XI, stated, ASME Code Interpretations are not incorporated into the Code of Federal Regulations and, therefore, the NRC is not bound by these interpretations. Although the NRC recognized ASME as the official interpreter of the ASME Code, Code Interpretations were not incorporated by reference into 10 CFR 50.55a and did not receive NRC approval.
Based on the above, the inspectors, in consultation with NRR, determined that Constellation misapplied Interpretation XI-1-13-25 given the background information leading to the issuance of the Interpretation. In addition, Constellation did not have an NRC-approved Code relief request to use the Interpretation as the basis to consider the surface areas of flued head components and welds at containment penetrations X-7A, X-7B, X-7C, X-7D, X-9A, X-9B, and X-8 as inaccessible due to the removable insulation covering the surface areas.
Notwithstanding, in consultation with NRR, the inspectors determined there is likely inconsistency in licensee approaches to define insulation as temporary or permanent and beyond this consideration, the NRC has broader safety and regulatory concerns involving the potential for some licensees to not perform in-service examination of containment welds for the life of their plant. The NRC's planned actions to address broader concerns are documented in ML23046A398 and involve in part, gathering further input from regional inspectors regarding what is necessary for these examinations and considering whether a condition on the ASME Code is necessary as part of agency 10 CFR 50.55a rulemaking for endorsing the 2023 Edition of the ASME Code.
Significance: The inspectors assessed the significance of this concern to establish that, in this instance, the concern if developed into a finding involves very low safety significance. Using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined a resulting finding would be of very low safety significance (Green) in accordance with Exhibit 3 because the finding did not represent an actual open pathway in the physical integrity of reactor containment. Also, the licensee's performance of 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," provided reasonable assurance for reactor containment integrity. Therefore, this issue is appropriate for the Very Low Safety Significance Issue Resolution Process.
Technical Assistance Request: No Technical Assistance Request was processed in support of this issue. This issue is being addressed by the NRC as documented in ML23046A398.
Corrective Action Reference: IR 04551206
Failure to Perform Preventive Maintenance on Raw Water Manual Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems
Green NCV 05000333/2022004-01 Open/Closed
None (NPP)71111.12 The inspectors identified a Green finding and associated non-cited violation (NCV) of Technical Specifications (TS) 5.4.1(a), Procedures, for failure to establish preventive maintenance in accordance with station procedures for a safety-related component.
Specifically, emergency service water manual valve 46ESW-12A was not receiving preventive maintenance. As result, a stem to disc separation occurred causing a loss of emergency service water supply to the A train emergency core cooling systems.
Description:
The emergency service water (ESW) system is a two-train system, each providing 100 percent cooling requirements for all safety-related equipment in the corresponding system train, including the emergency diesel generators. Upon loss of normal service water, the emergency service water system supplies the safety-related crescent area coolers. There are five coolers in each train, providing area room cooling to the emergency core cooling system pumps during a design basis loss of coolant accident and loss of offsite power.
On December 2, 2022, Constellation staff were performing ST-8Q, Testing of the Emergency Service Water System (IST), Revision 057 under Work Order 5263649. During the test, operators identified the west crescent unit coolers (66UC-A,C,E,G,I) were not receiving flow.
Operators suspended the test, entered Condition A of Technical Specification Limiting Condition for Operation (LCO) 3.7.2, Emergency Service Water System and Ultimate Heat Sink (UHS)," requiring restoration to Operable status within 7 days, entered Condition A of Technical Requirements Manual (TRM) specification 3.7.C, "Crescent Area Ventilation System," requiring restoration to Operable status within 7 days, and entered the issue into the corrective action program as issue report (IR) 4540603 and began investigation. On December 3, 2022, Constellation staff identified that manual supply 3-inch isolation valve, 46ESW-12A, had failed due to stem to disc separation. This resulted in isolation of the A ESW train supply to the five coolers in the west crescent. This was a loss of safety function to provide area room cooling to the A and C residual heat removal pump motors, A core spray pump motor, and the reactor core isolation cooling steam driven pump. Following repair of 46ESW-12A and successful unit cooler flow testing, Constellation exited the associated LCO and TRM requirements on December 4, 2022.
The inspectors identified during their review that 46ESW-12A was classified as non-critical, with no applicability for duty cycle, and run-to-maintenance with no preventive maintenance having ever been performed or scheduled. Contrary to this determination, Constellation procedure ER-AA-200-1001, Equipment Classification, Revision 4, dated January 18, 2018, 4, Component Classification Guidance and Examples, Step 24 provides guidance to assess manual valves. Step 24.D states, Manual valves that could result in unacceptable consequence due to stem to disc separation should be considered for a non-critical classification and an appropriate maintenance strategy developed.
In accordance with ER-AA-200, Preventive Maintenance Program, Revision 4, dated January 18, 2018, Step 4.1.7 states vulnerabilities are evaluatedto ensure appropriate strategies are developed to address the vulnerability, as required, and risk assessed if needed. In 2018, the station performed an equipment reliability and vulnerability review as directed in IR 4145473 on June 8, 2018. As a result of the stations review, on October 24, 2018, IR 4146063 assignment 26 was generated and actions completed following review of vulnerabilities for ESW. The review included ESW supply isolation and noted specific review for stem to disc separation on passive components where failure would block or restrict flow. Station personnel failed to identify the vulnerability associated with 46ESW-12A. As a result, no preventive maintenance was established, leading to stem to disc separation on December 3, 2022.
Corrective Actions: Constellation replaced the stem and disc of 46ESW-12A. In addition, the station wrote IR 4543877, to address the missed opportunity to review manual valves in accordance with ER-AA-200-1001. The station also wrote IR 4540709 to address the identified stem to disc separation and to drive actions to identify manual valves, and develop a ranking system to develop appropriate preventive maintenance.
Corrective Action References: IRs 4540709, 4543877
Performance Assessment:
Performance Deficiency: Failure to establish preventive maintenance in accordance with station procedures for a safety-related component.
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, emergency service water manual valve 46ESW-12A was not receiving preventive maintenance. As result, a stem to disc separation occurred, causing a loss of emergency service water supply to the A train emergency core cooling systems. This resulted in an entry to Condition A of Technical Specification 3.7.2, "Emergency Service Water System and Ultimate Heat Sink (UHS)," requiring restoration to Operable status within 7 days, and entry into Condition A of Technical Requirement Manual specification 3.7.C, Crescent Area Ventilation System, requiring restoration to Operable status within 7 days.
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 this finding to be of very low safety significance (Green) in accordance with Exhibit 2, because
- (1) it did not involve a deficiency affecting the design or qualification of a mitigating SSC that affected its operability or Probabilistic Risk Assessment (PRA)functionality;
- (2) it 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;
- (3) it did not represent a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time;
- (4) it did not represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
- (5) it did not represent a loss of a PRA system and/or function as defined in the Plant Risk Information Book or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- (6) it did not represent a loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with Constellation's 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. The failure to identify the vulnerability occurred outside the 3-year window.
Enforcement:
Violation: Constellation Technical Specification 5.4.1(a), Procedures, requires in part, that written procedures shall be established, implemented, and maintained covering the activities referenced in Regulatory Guide 1.33, Appendix A, November 1972. Regulatory Guide 1.33, Appendix A,Section I.1 states, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Regulatory Guide 1.33, Appendix A,Section I.2 states, in part, that preventive maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime.
Constellation procedure ER-AA-200, Preventive Maintenance Program, Revision 4, step 4.1.7 states vulnerabilities are evaluatedto ensure appropriate strategies are developed to address the vulnerability, as required, and risk assessed if needed. Step 4.1.1 states, component classifications and changes thereto are documented in the EStrategy Tool by the Strategic Engineer per ER-AA-200-1001. Procedure ER-AA-200-1001, Equipment Classification, Revision 4, Attachment 4, Component Classification Guidance and Examples, Step 24 provides guidance to assess manual valves. Step 24.D states, Manual valves that could result in unacceptable consequence due to stem to disc separation should be considered for a non-critical classification and an appropriate maintenance strategy developed.
Contrary to the above, from October 24, 2018 to December 3, 2022, the station failed to establish preventive maintenance in accordance with station procedures for a safety-related component. Specifically, emergency service water manual valve 46ESW-12A was not receiving preventive maintenance. As result, a stem to disc separation occurred causing a loss of emergency service water supply to the A train emergency core cooling systems.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Emergency Core Cooling Pump Switch Failures Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems
Green NCV 05000333/2022004-02 Open/Closed
[P.2] -
Evaluation 71111.13 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify a condition adverse to quality associated with main control room emergency core cooling system (ECCS) pump switches.
Specifically, Constellation failed to identify a potential common failure mechanism following three failures of ECCS main control room pump switches. As a result, Constellation did not properly classify significance following the third failure and although action items were established, a corrective action was not assigned in accordance with Constellation corrective action program requirements.
Description:
The ECCS at FitzPatrick includes four residual heat removal (RHR) pumps which provide low head safety injection, their associated four RHR service water pumps and two core spray pumps. Each of the RHR pumps and core spray pumps receive signals to automatically start during a loss of coolant accident due to high drywell (containment)pressure signal or low reactor water level. The pumps can also be started and stopped by operators using safety-related switches from the main control room during implementation of emergency operating procedures.
On November 30, 2021, while performing a quarterly surveillance test, the C RHR pump failed to start when the operator turned the switch to start. Constellation entered IR 04463794 into the corrective action program (CAP) and performed troubleshooting. The switch contacts were cleaned by maintenance, and the test was re-performed successfully. Issue report 04463794 was classified as Significance Level 4 and a work group evaluation (WGE) was assigned. Significance Level 4 is the lowest significance and only requires the discrepancy to be corrected. The WGE identified two potential causes. The first potential cause was associated with switch malfunction. The second potential cause was associated with operator error during switch operation. Actions to review the potential impact of the failure on other ECCS switches were closed because no cause was determined.
On March 1, 2022, while performing a quarterly surveillance test, the A core spray pump failed to start when the operator turned the switch to start. Constellation entered IR 04481502 into the corrective action program and performed troubleshooting. Troubleshooting revealed varying resistance readings when turning the switch. Constellation replaced the switch, and successfully re-performed the quarterly pump test. Following the A core spray pump switch failure, Constellation also scheduled and replaced the C RHR pump switch on October 7, 2022.
Issue report 04481502 was classified as Significance Level 4 and a WGE was assigned. The WGE identified 2 potential causes. The first potential cause was associated with switch malfunction. The second potential cause was associated with operator error during switch operation. Constellation sent the switch to Exelon PowerLabs for further investigation. The investigation did not identify an exact cause, based on successful operation in the lab. An action item to assess extent of condition and evaluate the need for additional preventive maintenance was closed because no cause was identified. Although not tracked by a CAP action, Constellation implemented standing order 25-018 allowing operators up to 5 pump start attempts using the control room switch and if unsuccessful, manual start attempts at the pump breaker.
On December 9, 2022, while performing a quarterly surveillance test, the B core spray pump failed to start when the operator turned the switch to start. The operator successfully started the pump on the third start attempt and completed the pump test successfully. Constellation entered IR 04541686 into the CAP and performed replacement of the switch. Constellation also expedited Exelon PowerLab investigation of the C RHR pump switch. On December 15, 2022, the investigation identified erratic resistance and sulfidation as the cause for the C RHR switch failure. Constellation screened IR 04541686 as a Significance Level 4 and investigation classification of D.
The inspectors reviewed PI-AA-120, Issue Identification and Screening Process," Revision 12 Step 4.6.6 and determined that Significance Level 3 was the correct classification. Level 3 Guidance example "ggg" states, Multiple examples of similar safety related equipment problems, where the equipment is operable, but has degraded, reflecting a potential common failure mechanism. The inspectors determined that degradation/sulfidation was identified as a potential failure mechanism for each switch during troubleshooting or PowerLabs investigation.
In addition, the inspectors reviewed PI-AA-125, Corrective Action Program (CAP)
Procedure, Revision 8. Section 4.5.2 states in part, Create a corrective action (CA) for any planned action necessary to restore a condition adverse to quality. The following guidance should be used to determine if the action is a CA. Actions that correct a Significance Level 1, 2, or 3 Condition. Contrary to the requirements of PI-AA-120 and PI-AA-125, Constellation did not properly classify and assign a CA for the repeat failure of an ECCS pump switch with a potential common failure mechanism.
Corrective Actions: Constellation replaced the three failed ECCS pump switches and instituted an operator standing order for ECCS pump switches. The standing order allows operators multiple attempts to manually start an ECCS pump from the control room or locally at the breaker.
Corrective Action References: IRs 04463794, 04481502, 04541686
Performance Assessment:
Performance Deficiency: Failure to identify a condition adverse to quality associated with main control room ECCS pump switches.
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, Constellation did not properly classify significance following the third failure and although action items were established, a corrective action was not assigned in accordance with Constellation CAP requirements. As a result, Constellation failed to identify a potential common failure mechanism following three failures of ECCS main control room pump switches.
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 this finding to be of very low safety significance (Green) in accordance with Exhibit 2, because it
- (1) did not involve a deficiency affecting the design or qualification of a mitigating SSC that affected its operability or PRA functionality;
- (2) 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;
- (3) did not represent a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time;
- (4) did not represent a loss of the PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
- (5) did not represent a loss of a PRA system and/or function as defined in the Plant Risk Information Book or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- (6) did not represent 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: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Constellation did not properly classify significance following the third failure and although action items were established, a corrective action was not assigned in accordance with Constellation corrective action program requirements.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.
Procedure PI-AA-120, Issue Identification and Screening Process," Revision 12, Step 4.6.6.3 directs the determination of Significance and Investigation Class of each CAP issue.
2 Significance Level 3 example "ggg" states, Multiple examples of similar safety related equipment problems, where the equipment is operable, but has degraded, reflecting a potential common failure mechanism.
Procedure PI-AA-125, Corrective Action Program (CAP) Procedure, Revision 8, Section 4.5.2 states, in part, Create a corrective action (CA) for any planned action necessary to restore a condition adverse to quality. The following guidance should be used to determine if the action is a CA. Actions that correct a Significance Level 1, 2, or 3 Condition.
Contrary to the requirements of PI-AA-120 and PI-AA-125, following the third ECCS pump switch failure on December 9, 2022, Constellation did not properly classify and assign a corrective action for the repeat failure of an ECCS pump switch with a potential common failure mechanism.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Lead Shielding Not Analyzed in Accordance with Station Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity
Green NCV 05000333/2022004-03 Open/Closed
None (NPP)71111.15 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for Constellation's failure to evaluate shielding in accordance with station procedures. Specifically, the shielding was in contact with reactor building closed loop cooling piping which is seismic class I and serves as part of the containment boundary.
Description:
On October 6, 2022, during a plant walkdown, the inspectors noted that permanent shielding was in contact with four-inch piping of the reactor building closed loop cooling system. The piping and nearby valve are classified as seismic class I and provide cooling to the recirculation pumps during power operation and serve as the containment isolation boundary during accident conditions. The shielding was installed under shielding package 1992-090. The inspectors questioned the acceptability of the installed configuration and the operability of the reactor building closed loop system including the containment isolation function of this piping. The engineering department documented the issue in IR 04539840.
Procedure RP-ALARA-01.04, "Shielding," Revision 8, effective June 6, 2006 required all existing long term shielding to be evaluated by engineering and converted to permanent shielding. The engineering evaluation could not be located or provided for shielding package 1992-090. The shielding was modified to remove the contact with the safety related piping and the weight of the shielding was reduced. Constellation engineering analyzed the corrected configuration and determined that it was acceptable.
Corrective Actions: The licensee entered the issue into their corrective action program. The shielding quantity was reduced and contact was eliminated to the safety-related pipe.
Engineering performed an evaluation of the new acceptable configuration.
Corrective Action References: 04539840
Performance Assessment:
Performance Deficiency: The licensee's failure to evaluate the shielding in contact with safety-related components per applicable procedures is a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control 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. As a result, Constellation modified the field configuration to reduce the shielding quantity and eliminate contact with the safety-related pipe to eliminate unevaluated seismic induced pipe loads.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Specifically, the finding screened to Green (very low safety significance) in accordance with Exhibit 3 because the finding did not represent an actual open pathway in the physical integrity of reactor containment, failure of containment isolation system, failure of containment pressure control equipment, failure of containment heat removal components, or an actual reduction of hydrogen igniters.
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. The shielding was installed in 1992 and engineering analysis was required in 2006.
Enforcement:
Violation: Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be described by documented instructions or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions or drawings. Procedure RP-ALARA-01.04, "Shielding,"
Revision 8, required all existing long term shielding to be evaluated by engineering and converted to permanent shielding. Contrary to the requirement, from 1992 through 2022 shielding was in contact with safety related equipment without an engineering evaluation as required by station procedures.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Licensee-Identified Non-Cited Violation 71111.15 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be described by documented instructions or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions or drawings. Procedure MA-AA-716-010, "Maintenance Planning," Section 4.20.2 requires that an engineering change request (ECR) be generated when a deviation from the original plant design is identified during the planning process. The ECR is the process used by engineering to evaluate and document the acceptability of these changes.
Contrary to above, on December 4, 2022, Constellation failed to generate the necessary ECR and evaluate the use of non-qualified parts in the safety-related emergency service water system. Specifically, although the internal components for 46ESW-12A were identified as non-safety they were planned, released, and installed without engineering review or approval.
Significance/Severity: Green. The inspectors determined this issue to be Green in accordance with IMC 0609 Appendix A, Exhibit 2 because the finding is a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintains its operability.
Corrective Action Reference: IR 04541193
Minor Performance Deficiency 71152A The inspectors reviewed Constellations non-corrective action program IR 04443511 and associated department review, briefing sheets, performance analysis of corrective action program assignments, training presentations, and audits. The inspectors also interviewed responsible Constellation staff regarding this IR, which identified the potential negative trend in corrective action program completion documentation such as incomplete or lack of closure justification.
Constellation performed a department review of the issue, which was documented in IR 04443511. As a result of the review, Constellation identified a negative trend in their corrective action program existed, regarding completion documentation, such as incomplete or lack of closure justification. This was identified from internal document reviews, NRC Problem Identification and Resolution inspections, and Constellation Nuclear Oversight Corrective Action Program Audit. As part of the department review, Constellation identified non-conservative categorization of NRC and Nuclear Oversight identified issues, citing 14 examples.
Constellation identified that when an NRC issue is identified it is captured by an IR and categorized as either corrective action program or non-corrective action program based on the significance and actual impact on the plant. By categorizing IRs as non-corrective action program as opposed to corrective action program, layers of oversight such as management review committee review, are removed. Furthermore, based on a review at FitzPatrick, Constellation found they categorize more NRC identified IRs as non-corrective action program items than the rest of the fleet. The inspectors' review of corresponding corrective action program procedures note that the categorization and classification of an IR is completed by the site, given specific details of the issue.
The inspectors reviewed the non-corrective action program IR, department review of the issue, and subsequent site action assignments. The inspectors noted that IR 04443511 was not entered in the corrective action program, it was categorized as a non-corrective action program document. A review of the applicable Constellation procedure PI-AA-120, Issue Identification and Screening Process. Revision 12, details steps that indicate that IR 04443511 should have been entered into the corrective action program. Specifically, 2 of PI-AA-120 states in part that, Issues where there is a potential concern or risk for plant and cyber/security equipment, nuclear regulatory, and accredited training process issues and QA [quality assurance] program non-compliances should remain in CAP
[corrective action program]. The inspectors concluded that inadequately completing corrective action program actions as stated in IR 04443511 is an issue where there is a potential concern or risk for nuclear regulatory and QA program non-compliance.
Consequently, IR 04443511 should have been entered into the corrective action program as opposed to the non-corrective action program.
The inspectors additionally identified that procedure PI-AA-101-1001, Performance Monitoring and Analysis Manual Revision 3, Section 4.2 Trend Analysis, paragraph 4.2.3 states, in part, Trend IRs should be evaluated utilizing the WGE (work group evaluation)analysis method in the CR/NCAP Processing web tool unless the Cause, Extent of Condition and Actions to address are documented in the Issue Report. The inspectors note that trend IR 04443511 did not perform a work group evaluation, which is a corrective action program action. Constellation did complete a department review of the issue, which was documented in the IR and presented the conclusions to the management review committee.
Also, the absence of a work group evaluation was essentially completed in part by the department review. However, the inspectors note that Constellation had established procedural steps and an opportunity by procedure to enter the issue of concern into the corrective action program. Had the issue been in the corrective action program by procedure, a work group evaluation would have been performed and corrective action program assignment assigned to Constellation staff consisting of additional reviews and established processes.
Screening: The inspectors determined the performance deficiency was minor. The inspectors determined that although not greater than minor significance, the placement of IR 04443511 into the non-corrective action program as opposed to the corrective action program was a performance deficiency and Constellation's actions were not commensurate with the safety significance of the issue. Specifically, the inspectors independently evaluated the deficiencies noted above for significance in accordance with the guidance in IMC 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues. The inspectors determined that none of the conditions were deficiencies of greater than minor significance and therefore are not subject to enforcement action in accordance with the NRCs enforcement policy.
Observation: FitzPatrick 1R25 Deferral of N-2F-SE Nozzle to Safe End UT Exam 71152A The inspectors reviewed Constellation performance associated with IR 04450965, CPO Oversight Actions, J1R25 Deferral of N-2F-SE UT Exam, initiated on October 1, 2021. The inspectors reviewed nondestructive evaluation (NDE) reports, engineering evaluations, examination procedures, ASME Section XI Code requirements, BWRVIP [Boiling Water Reactor Vessel and Internals Program]-75A reports and interviewed Constellation staff regarding this IR. The inspectors reviewed Constellations technical standards applicable to weld examination, evaluation and documentation and their corrective action process procedures.
Issue report 04450965 documented Constellation staff views that recirculation nozzle inlet dissimilar metal nozzle to safe end weld N-2F-SE should be volumetrically examined during refueling outage J1R25 in fall 2022. FitzPatrick staff described in their IR that a flaw was identified in the weld when examined in January 2017 using manual, non-encoded phased array ultrasonic testing. The flaw was characterized as fabrication induced (i.e., from original construction) axially oriented and totally contained within the butter and weld material. The exam record showed the flaw to be acceptable per ASME Section XI Table IWB-3514-2, Allowable Planar Flaws.
Constellation staff at the FitzPatrick plant identified in IR 04450965 that external subject matter experts subsequently recommended preservice and prior in-service examination records be reviewed. The IR indicated this review was completed with the result that there was no evidence observed of a flaw at this location, lending support to a service-induced flaw. Constellation staff further documented in their IR that a flaw present over multiple weld passes and in both the weld and weld butter materials was unlikely at the fabrication stage.
Constellation staff further identified in their IR that industry subject matter experts recommended the weld be re-examined at the next refueling outage. However, the exam was not conducted during the refueling outage in 2020. Instead, in November 2020, Constellation performed an engineering evaluation to assess whether the flaw in the weld N-2F-SE was acceptable for continued operation with the assumption that the flaw was service-induced and connected to the inner diameter of the pipe. The evaluation predicted that the flaw would exceed applicable ASME Code Section XI acceptance criteria after approximately 12.72 years from the last inspection time referenced (2017) which would be in 2029. The IR stated the examination was subsequently descoped from the fall 2022 outage and scoped into the fall 2026 outage. Constellation documented their technical basis and conclusions regarding plant risk associated with this deferral in Risk Classification Manager entry JAF-1-2021-0273. The entry restated the history of decisions for this problem, the evaluation results, and the risk being an axially initiated leak at this location resulting in plant shutdown based upon increased unidentified leak rate and concluded that the risk rank and likelihood were both low.
The inspectors considered that there was significant uncertainty in Constellations conclusion of low likelihood categorization. While the inspectors' review of Constellations 2020 evaluation found that the methods, material property and operational parameter inputs were technically supported, there was significant uncertainty introduced as to the extent of the flaw as the flaw tips were not identified and plotted in the exam record in 2017. Additionally, the inspectors noted (as did FitzPatrick staff in their IR) that the N-2F-SE examination data sheet dated February 1, 2017, did not provide typical characterization detail for a reviewer to independently assess the flaw. Because the flaw tips are used to identify the beginning and end of the crack, it is one of the primary inputs to the evaluation to determine how long it will take for the flaw to exceed ASME structural acceptance criteria. The inspectors considered that using inaccurate inputs may lead to non-conservative results. Notwithstanding, the inspectors' review of the applicable procedure in effect at that time did not identify a standard that was not met.
Furthermore, the inspectors noted that conclusions reached based on the data gathered were in question in light of prior exam results. The inspectors also noted that weld N-2F-SE was examined in 2017 as an expansion sample based on finding intergranular stress corrosion cracking in residual heat removal weld 24-10-130, which was repaired by weld overlay. As a result, the inspectors considered service-induced degradation mechanisms are possible in this weld. Therefore, the inspectors considered Constellation's conclusions that weld N-2F-SE is suitable for continued service to the fall 2026 refueling outage involved significant uncertainty not reflected in their categorization as low likelihood.
Minor Performance Deficiency 71152S The inspectors evaluated a sample of issues and events that occurred over the third and fourth quarters of 2022. The evaluation did not reveal any new trends that could indicate a more significant safety issue. The inspectors determined that, in most cases, the issues were appropriately evaluated by Constellation staff for potential trends at a low threshold, and resolved within the scope of the CAP.
The inspectors reviewed a CAP evaluation of an equipment reliability trend identified by the NRC in IR 04473958. Specifically, Constellation staff determined there were no common issues. The inspectors determined the assessment did not include potential organizational or equipment reliability common issues such as preventive maintenance strategy for components being effective maintenance practices, adequacy of maintenance procedures, or the effect of aging of equipment in accordance with PI-AA-125-1006, "Investigation Techniques," Revision 6, Attachment 15.
Following the inspectors' observations in the second quarter of 2022, the station performed a review of organizational and behavioral trends. The station identified that operational-type procedures do not specify declaration of mitigating systems performance indicator (MSPI)impact or safety system performance metrics impact as unavailable or available. This included not logging start and stops for equipment being cycled. Corrective actions were completed in December 2022 to identify and flag MSPI and safety system performance metrics in the plan of the day package, work week schedule and planning packages for better awareness.
The inspectors determined that although there are new actions to address the areas of awareness, the station still has not addressed continued challenges with equipment failures.
These include:
- A low pressure coolant injection battery cell 171 being outside of parameter acceptance criteria
- B core spray pump switch failure
- Refuel bridge failures during the J1R25 refuel and maintenance outage
- A inboard main steam isolation (LPCI) valve failing to close with an isolation signal
- G safety relief valve main control room switch failure
- Reactor mode switch failure
- 600V supply breaker failures: 71-11502, 71-11302, and 71-12302
- Valve maintenance issues that included hammer events of an LPCI injection valve 10MOV-25A and a main steam drain valve, 29MOV-74, and motor damage for a reactor water recirculation isolation valve 02-2MOV-53A
- 46ESW-12A, manual valve stem to disc separation in emergency service water supply
Constellation generated IR 04546406 to perform a common cause analysis. Based on the overall results of the semi-annual trend review, the inspectors determined that Constellation had generally identified adverse trends before they could become more significant safety problems.
Screening: The inspectors determined the performance deficiency was minor. The inspectors independently evaluated the deficiencies noted above for significance in accordance with the guidance in IMC 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues, and determined them to be minor. The inspectors also identified findings of more than minor significance associated with the 46ESW-12A valve failure and the B core spray switch failure which are documented in this report.
Observation: LER 05000333/2022-001-00, Exhaust Drain Pot Line Filled with Water Up to HPCI Turbine Due to Relay Failure 71153 The inspectors determined that the retraction of this LER (EN#55871) appeared reasonable based on Constellations evaluation of the impact on the HPCI system due to the accumulated water in the HPCI turbine. Specifically, a technical evaluation of the condition was performed and concluded that the HPCI system would have remained operable. This LER is Closed.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On February 1, 2023, the inspectors presented the integrated inspection results to Garrick Olson, Plant Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
04506624
Procedures
OP-AA-108-111-
1001
Severe Weather and Natural Disaster Guidelines
Seasonal Readiness
WC-JF-107
Seasonal Readiness T&RM for JAF
Work Orders
05193435
Drawings
Decay Heat Removal Flow Diagram
Decay Heat Removal Flow Diagram
Miscellaneous
Maintenance Rule
Basis Document
System 032
Decay Heat Removal System
Procedures
Decay Heat Removal System
Procedures
Containment Atmospheric Dilution System
Fire Plans
Relay Room/ Elev. 286' Fire Area/Zone VII/RR-1
Main Control Room & Control Room HVAC Equipment
Rooms/Elev. 300'
Calculations
JAF-CALC-16-
00005
Reactor Building Water Spray Curtin Internal Flooding
Analysis
Miscellaneous
JAF-NE-09-00001 James A. FitzPatrick Nuclear Power Station Probabilistic
Safety Assessment Internal Flooding
Work Orders
04971965
262010
71111.08G Corrective Action
Documents
04051628
04450965
Drawings
DWG ISI-
IWE_004
Typical Containment Penetrations
Revision 1
Engineering
Changes
DCR 02.120
Creation of IWE (MC Component) Inspection Drawings
06/03/2002
Engineering
Evaluations
2001089.301P
Flaw Evaluation of N-2F-SE Nozzle to Safe End Weld NDE
Indication
11/22/2020
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Constellation email, "ASME XI Requirements for Insulation
Removal" (ML23004A020)
09/30/2022
Duke Energy Letter, "ASME Code,Section XI, IWE-2500,
1992 Edition with the 1992 Addenda through the 2013
edition"
11/13/2013
JAF-1-2021-0273
Risk Assessment: Document Technical Basis and Plant Risk
associated with Deferral of N-2F-SE Examination from
J1R25(2022)
05/03/2022
XI-1-13-25
ASME Interpretation Detail "Inquiry on IWE-2500 related to
accessibility for examination"
03/07/2015
NDE Reports
2-UT-019
N-2F-SE/Recirculation System
10/16/2002
08UT055
Examination Summary Sheet N-2F-SE Safe End to Nozzle
09/30/2008
JAF-27-N-2F-SE
Nozzle to Safe End N-2F-SE
2/01/2017
Work Orders
05133066-14
IWE Containment Exams
09/25/2022
71111.11Q Miscellaneous
FPC26-SU-
SEQ1.0-A2
Control Rod Move Sheet NF-AB-720 F-2
Procedures
Startup and Shutdown Procedure
137
Startup and Shutdown Procedure
138
Corrective Action
Documents
04540603
04540676
Procedures
Maintenance Planning
MA-AA-716-010-
1000
Passport Work Planning Manual
Suppression Chamber and Drywell Deterioration Inspection
Corrective Action
Documents
04540603
04540676
04541686
Miscellaneous
James A. FitzPatrick Nuclear Power Plant Probabalistic Risk
Assessment Core Spray System Notebook
Procedures
Refueling Water Level Control
Integrated Risk Management
Drain Down of Reactor Cavity to Torus Using RHR Loop B in 4
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
04524479
04526683
04527635
04528655
04528757
04529539
04529651
04540510
Corrective Action
Documents
Resulting from
Inspection
04539840
Engineering
Changes
637547
Through-Wall Leak on 'B' Emergency Service Water (ESW)
8-inch Line
637732
Defeat Overtravel Annunciator for Control Rod 14-39
638028
Re-Evaluate Load of Lead Shielding Blankets Allowed to be
Placed on 4-Inch Diameter Drain Lines located on Top of the
Drywell Mezzanine
Miscellaneous
01355772
Purchase Order
Procedures
Maintenance Specification: Installation and Control of
Temporary Shielding
ISI-FM-15B
ISI Flow Diagram Reactor Bldg Cooling Water System 15
Guidelines for Installation and Control of Spot Shielding
Control Rod Coupling Integrity Test
Work Orders
04714594
Corrective Action
Documents
04526744
04528757
Engineering
Changes
635090
637732
Defeat Overtravel Annunciator for Control Rod 14-39
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
Uncoupled Control Rod
Rod Overtravel
Control Rod Exercise, Timing, and Stall Flow Test
Control Rod Coupling Integrity Test
RCIC Flow Rate and Inservice Test (IST)
Work Orders
05085379
05185217
299941
Calculations
MIDACALC
Results 10MOV-
25A
AC Motor Operated GL96-05 Gate Valve Calculation Results
Corrective Action
Documents
04525932
04527843
Drawings 1.83-39, Sheet1
Elementary Diagram Auto Depressurization System
Miscellaneous
CR-JAF-2005-
01956
CR-JAF-2008-
03408
Procedures
1.75-70
Elementary Diagram Feedwater Control System
Elementary Diagram RFPT Control Sheet 1
Flow Diagram Standby Liquid Control System 11
EDG Governor Tuning Procedure
Standby Liquid Control Explosive Valves (IST)
RPV High Water Level Trip of Feedwater Pump Turbine and
Main Turbine Logic System Functional Test
RPV System Leakage Test
HPCI Quick-Start, Inservice, and Transient Monitoring Test
(IST)
Standby Liquid Control Recirculation, Injection Test (IST,
ISI)
EDG A and C Load Sequencing Test and 4 kV Emergency
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Power System Voltage Relays Instrument Functional Test
Work Orders
04883485
04897826
05081873
05094548
05117640-01
05117867
236193
236193
293252
297937
Corrective Action
Documents
Resulting from
Inspection
04526637
Procedures
Flow Diagram Fuel Pool Cooling & Cleanup System 19
Fuel Pool Cooling and Clean-up System
Startup and Shutdown Procedure
138
Work Orders
22579-09
Corrective Action
Documents
04355085
04406552
04527251
04527602
Procedures
Diverse and Flexible Coping Strategies (FLEX) Spent Fuel
Pool Instrumentation (SFPI), and Hardened Containment
Vent System (HCVS) Program Document
CC-JF-118
Site Implementation of Diverse and Flexible Coping
Strategies (FLEX) and Spent Fuel Pool Instrumentation
Program
Floor Drain Flow Test - Plant Shutdown
FSG-001
Initial Assessment and FLEX Equipment Staging
Diesel Fire Pump Engine 76P-4(ENG)
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
RHR Loop A Containment Spray Headers and Nozzles Air
Test (ISI)
East Diesel Fire Pump Performance Test
Work Orders
05061257
05117649-01
05117780-01
05171528
Procedures
N-JF-OPS-
LORTEVAL-43A
11/11/2022
71151
Corrective Action
Documents
04461845
04496616
Drawings
Flow Diagram Residual Heat Removal
Engineering
Changes
636436
Procedures
ER-AA 600-1047
Mitigating Systems Performance Index Basis Document
Mitigating System Performance Index Data Acquisition &
Reporting
Monthly Data Elements for NRC ROP Indicator Safety
System Functional
71152A
Corrective Action
Documents
04473958
71152S
Corrective Action
Documents
04524574
04524581
04525932
04527459
04527843
04527898
04527980
04528036
04528466
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
04528629
04535796
04535798
04535802
04535820
04535825
Calculations
JAF-CALC-06-
00030
JAFNPP Structural Qualification of HPCI Turbine Steam
Exhaust from Turbine to Sparger in Torus
Corrective Action
Documents
04496616
04534560
04535530
04538719
Elementary Diagram High-Pressure Coolant Injection
System
Engineering
Evaluations
007N1485
Project Task Report, HPCI Flooding/Water Hammer
Concerns
Procedures
HPCI VLV or MTR Overload or Cntrl Pwr Loss
Work Orders
05309025