IR 05000333/2023002

Integrated Inspection Report 05000333/2023002

ML23219A114
Person / Time
Site:	FitzPatrick
Issue date:	08/07/2023
From:	Erin Carfang NRC/RGN-I/DORS
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
References
IR 2023002
Download: ML23219A114 (1)

Text

August 7, 2023

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - INTEGRATED INSPECTION REPORT 05000333/2023002

Dear David Rhoades:

On June 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at James A. FitzPatrick Nuclear Power Plant. On July 26, 2023, the NRC inspectors discussed the results of this inspection with Timothy Peter, Site Vice President, 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. These findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at 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, Digitally signed by Erin E.

Erin E. Carfang Date: 2023.08.07 Carfang 15:26:50 -04'00'

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: DPR-59 Report Number: 05000333/2023002 Enterprise Identifier: I-2023-002-0037 Licensee: Constellation Energy Generation, LLC Facility: James A. FitzPatrick Nuclear Power Plant Location: Oswego, NY Inspection Dates: April 1, 2023 to June 30, 2023 Inspectors: E. Miller, Senior Resident Inspector B. Sienel, Resident Inspector P. Cataldo, Senior Reactor Inspector J. England, Senior Construction Inspector S. Haney, Senior Resident Inspector T. Hedigan, Operations Engineer S. McCarver, Physical Security Inspector P. Ott, Operations Engineer B. Pinson, Senior Reactor Inspector D. Werkheiser, Senior Reactor Analyst Approved By: Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at 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.

List of Findings and Violations

Repetitive Failure to Follow Scaffolding Procedural Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.1] - 71152A Systems NCV 05000333/2023002-01 Identification Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, "Instructions,

Procedures and Drawings," because Constellation personnel did not adequately accomplish scaffold construction, inspection, and approval in the 'A' and 'C' residual heat removal service water (RHRSW) pump room in accordance with approved scaffold control procedures. Specifically, during walkdown inspectors identified scaffolding in contact with safety-related equipment without adequate evaluation.

Failure to Provide Adequate Work Instructions in Safety-Related Breaker Maintenance Procedures Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.7] - 71152A Systems NCV 05000333/2023002-02 Documentation Open/Closed A self-revealed Green finding and associated NCV of 10 CFR Part 50, Appendix B Criterion V,

Instructions, Procedures, and Drawings, was identified for failure to provide adequate work instructions in the maintenance procedure used to service and inspect safety-related 4160 volt (V) breakers. Specifically, maintenance procedures MP-054.01 and MP-054.03, used for servicing the closing coil assembly in 4160 V breakers, did not provide adequate guidance to ensure proper fit and alignment of required fiber spacers. The incorrect installation of the spacers resulted in misalignment of the closing coil sleeve which caused the associated armature to bind against the spacer and the end of the sleeve, increasing the amount of force required to close the breaker, and resulting in the failure of breaker 71-10504 to close on demand on April 12, 2023. Additionally, breaker maintenance procedures do not include specific guidance to ensure breaker to cubicle alignment is performed and maintained, which could allow for breaker shifting to occur over time, and lead to interference issues between the breaker and the cubicle.

Incomplete Mode Switch Position Verification Causes Group 1 Containment Isolation Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.11] - 71153 NCV 05000333/2023002-03 Challenge the Open/Closed Unknown A self-revealed Green finding and associated NCV of Technical Specification (TS) 5.4.1(a),

Procedures, was identified because the reactor shutdown procedure did not contain adequate verification steps for Mode Switch repositioning. As a result, when operators failed to verify all indications for the Mode Switch following repositioning on September 26, 2022, a reactor scram and Group 1 containment isolation occurred.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000333/2022-002-00 Mode Switch Failed to 71153 Closed Bypass Low Main Steam Line in Mode 2 Resulting in MSIV and RPS System Actuation LER 05000333/2022-002-01 Mode Switch Failed to 71153 Closed Bypass Low Main Steam Line in Mode 2 Resulting in MSIV and RPS System Actuation

PLANT STATUS

FitzPatrick began the inspection period at rated thermal power. On May 19, 2023, the unit was downpowered to 73 percent to perform turbine valve testing, control rod scram time testing and a control rod pattern adjustment. The unit was returned to rated thermal power on May 20, 2023. On May 21, 2023, the unit was downpowered to 85 percent to perform a control rod pattern adjustment. The unit was returned to rated thermal power on May 22, 2023. On June 28, 2023, the unit was downpowered to 75 percent to perform a condenser tube repair. The unit was returned to rated thermal power following repair on June 29, 2023 and remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

71111.01 - 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 warm temperatures for the 115 kilovolt and 345 kilovolt switchyard on June 28, 2023.

71111.04 - Equipment Alignment

Partial Walkdown (IP Section 03.01) (3 Samples)

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

(1) Standby liquid control system during the week of April 3, 2023

(2) Reactor core isolation cooling system on May 2, 2023

(3) 125 volts direct current temporary station battery charger 71BC-9 during 'A' station battery charger planned maintenance on May 9, 2023

Complete Walkdown (IP Section 03.02) (1 Sample)

(1) Emergency diesel generator (EDG) alignment during the week of April 17, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection (IP Section 03.01) (6 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) West cable tunnel, fire area/zone II/CT-2, on April 4, 2023

(2) Turbine building north, elevation 252', fire area/zone IE/TB-1, on April 20, 2023

(3) Security secondary alarm station, fire area/zone YARD, on April 21, 2023

(4) Standby gas filter room fire area/zone XX/SG-1 on April 21, 2023

(5) EDG spaces, fire area/zones V/EG-1, EG-2, and EG-5, on May 1, 2023

(6) Screenwell pump rooms, fire area/zones XII/SP-1 and SP-2 on May 4, 2023

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

(1) The inspectors evaluated fire brigade performance on April 4, 2023.

71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance

Requalification Examination Results (IP Section 03.03) (1 Sample)

(1) The inspectors reviewed and evaluated the licensed operator requalification exam results for the annual operating exam and biennial written exam completed on May 11, 2023.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(1 Sample)

(1) The inspectors observed operations personnel during the 'B' reactor protection system motor generator restoration on April 21, 2023; the 'A' reactor water recirculation speed controller failure to respond on May 4, 2023; and a reactor core isolation cooling surveillance on June 6, 2023.

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

(1) The inspectors observed a simulator evaluation that included a failure of a control rod drive pump, a failure of a 600 volt power supply, a loss of all feedwater, and loss of the 600 volt 11502 bus on April 11, 2023.

(2) The inspectors observed a simulator evaluation that included a condensate booster pump seal failure, an inadvertent high pressure coolant injection system start, and a failure of the containment isolation system on June 14, 2023.

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) Main control room switches for emergency core cooling systems on April 3, 2023

(2) 'A' residual heat removal pump start discharge pressure automatic depressurization system permissive relay failure on April 28, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management (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) Emergent maintenance following failure of the 'A' and 'C' emergency diesel tie breaker 71-10504 on April 12, 2023

(2) Elevated risk during 71T-2 115 kilovolt to 4 kilovolt offsite power supply transformer planned maintenance on May 22, 2023

(3) Elevated risk during high pressure coolant injection system planned maintenance on May 30, 2023

(4) Elevated risk during 'B' core spray planned maintenance on June 13, 2023

(5) Emergent downpower due to unplanned main condenser B1 tube leakage on June 28, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)

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

(1) Offgas charcoal following firefighting foam intrusion to the reactor coolant system on April 5, 2023

(2) Control rod 34-51 high temperature on April 16, 2023

(3) Evaluation of excess cables found on duct in reactor building east crescent on April 19, 2023

(4) 'B' and 'D' EDG fuel oil storage tank level on April 25, 2023

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

(1 Sample)

The inspectors evaluated the following temporary or permanent modifications:

(1) Temporary Modification: Scaffold 05290827 erected for inspection of north safety pump room fan 73FN-3A on May 22, 2023

71111.24 - Testing and Maintenance of Equipment Important to Risk

The inspectors evaluated the following testing and maintenance activities to verify system

operability and/or functionality: Post-Maintenance Testing (PMT) (IP Section 03.01)

(1) 'A' and 'C' emergency diesel tie breaker 71-10504 following failure to close during testing on April 12, 2023

(2) 'B' control room chiller, 70RWC-2B, following planned maintenance on May 4, 2023

(3) 'A' reactor water recirculation speed control and actuator following troubleshooting on May 11, 2023

(4) 71-10502 EDG 'A' feed to 10500 4 kilovolt bus following extent-of-condition inspection and closing coil spacer repair on May 15, 2023

(5) 'A' and 'C' EDG following 'C' EDG planned maintenance on May 18, 2023

(6) High pressure coolant injection system following planned maintenance on May 31, 2023

(7) 'B' core spray system following planned maintenance on June 13, 2023

Surveillance Testing (IP Section 03.01) (2 Samples)

(1) ST-9AB, EDG System B Fuel/Lube Oil Monthly Test, on April 26, 2023

(2) ISP-22-2, High-Pressure Coolant Injection (HPCI) System Low Flow Bypass Valve

Instruments, on May 4, 2023 Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) ST-2AM, residual heat removal (RHR) Loop B Quarterly Operability Test, on May 31, 2023

Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)

(1) Full load test of 200 kilowatt diesel FLEX-DG1 and FLEX-DG2 on June 8, 2023

71114.06 - Drill Evaluation

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

(1 Sample)

(1) The inspectors evaluated the conduct of a routine FitzPatrick emergency response drill that involved an earthquake, a failure of the reactor to scram, and a failure of containment to isolate on May 16,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

BI01: Reactor Coolant System (RCS) Specific Activity (IP Section 02.10) ===

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

BI02: RCS Leak Rate (IP Section 02.11) (1 Sample)

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

71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)

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

(1) Issue Report (IR) 04524574 - Incomplete Mode Switch Position Verification Causes Group 1 Containment Isolation and Reactor Scram

(2) IRs 04527980, 04528860, 04526744, 04525955, 04525964, 04527898, and

===04527981 - Outage-Related Valve Deficiencies

(3) IRs 04550060 and 04669409 - Review of 4 kilovolt Breaker Failures 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 05000333/2022-002-00 and LER 05000333/2022-002-01, Mode Switch Failed to Bypass Low Main Steam Line in Mode 2 Resulting in Main Steam Isolation Valves (MSIV) and Reactor Protection System (RPS) Actuation (ADAMS Accession Nos.

ML22327A241 and ML23031A225, respectively). The inspectors reviewed both LER submittals. The inspection conclusions associated with these LERs are documented in this report under Inspection Results. These LERs are Closed.

Personnel Performance (IP Section 03.03) (2 Samples)

(1) Earthquake and K9 geomagnetic disturbance on April 24, 2023

(2) 'A' reactor protection system motor generator output breaker trip on overvoltage on

May 25, 2023 Reporting (IP Section 03.05) (1 Sample)

(1) 70TS-109B control room temperature switch failure on April 6,

INSPECTION RESULTS

Repetitive Failure to Follow Scaffolding Procedural Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.1] - 71152A Systems NCV 05000333/2023002-01 Identification Open/Closed The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," because Constellation personnel did not adequately accomplish scaffold construction, inspection, and approval in the 'A' and 'C' residual heat removal service water (RHRSW) pump room in accordance with approved scaffold control procedures. Specifically, during walkdown inspectors identified scaffolding in contact with safety-related equipment without adequate evaluation.

Description:

In May 2023, during routine walkdowns, the inspectors identified scaffolding that was not constructed in accordance with Constellation procedures. Specifically, the scaffolding was in contact with the RHRSW strainer outlet flange and the emergency pump room exhaust fan housing (73FN-3A). On May 22, 2023, Constellation documented these issues in IR 04679899.

The inspectors identified that Constellation constructed and approved the scaffolding in the

'A' and 'C' RHRSW pump room contrary to the scaffold control and installation procedures. Specifically, the inspectors noted the following procedural requirements:

• MA-AA-716-025, "Scaffold Installation and Removal Request Process," Revision 19, requires that all braced scaffold members maintain a minimum horizontal and vertical separation of 2 inches or greater from safety-related equipment. Additionally, Section 4.1.4.3 directs Operations to determine if post-erection or removal inspections by Operations are warranted (i.e., scaffold location in close proximity to safety-related equipment).

• MA-AA-716-025-F-6, Operations Post Installation Walkdown Checklist, Revision 0, directs a review to determine if the scaffold maintains minimum spacing requirements from safety-related SSCs. This checklist was not performed for this installation.

• MA-AA-796-024-F-01, "Scaffold/Vertical Barrier Inspection Check List," Revision 0, Steps 15 and 16, verifies that the scaffold is not supported by, in contact with or connected to safety-related equipment, and that all seismic clearances are maintained.

The inspectors determined that contrary to these procedural requirements, Constellation constructed, inspected, and approved the scaffolding in the A and C RHRSW room although it did not meet procedural requirements, was in contact with the 'A' and 'C' RHRSW strainer and emergency pump room ventilation fan housing, and did not have an evaluation to allow this configuration.

Corrective Actions: Constellation performed a Human Performance Event Review Board, corrected the subject scaffolding, performed extent-of-condition walkdowns of all accessible scaffolding in the plant, and reviewed the scaffolding procedural requirements with all staff. Constellation also implemented procedure MA-JF-796-024-F-1, "Scaffold/Vertical Barrier Inspection Checklist" to include requirements for 2-inch clearance from safety-related equipment in future scaffold builds and identify areas requiring an Operations walkdown after scaffold construction.

Corrective Action References: IR 04679899

Performance Assessment:

Performance Deficiency: Scaffold was not constructed and evaluated in accordance with procedure prior to use.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control 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. The inspectors also noted that the performance deficiency was similar to Example 4.a of Inspection Manual Chapter (IMC) 0612, Appendix E. Specifically, each issue did not have an Engineering or Operations evaluation to assess seismic impact of the scaffolds in contact with safety-related equipment. The attachment of the scaffold to the safety-related support of fan 73FN-3A and the 'A' and 'C' RHRSW strainer rendered them subject to seismic induced loads that had not been considered in the original analysis and increased the probability of fan failure during accident mitigation.

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 did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g.,

seismic snubbers, flooding barriers, tornado doors) for greater than 14 days.

Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. The inspectors determined that Constellation personnel did not adequately construct scaffolding in safety-related structures in accordance with approved scaffold control procedures, and subsequent inspections by Operations, Maintenance and Engineering personnel did not recognize or correct this adverse condition, despite numerous similar issues recently entered into the corrective action program.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings states that, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Procedure MA-AA-716-025, "Scaffold Installation and Removal Request Process," Revision 19, requires that all braced scaffold members maintain a minimum horizontal and vertical separation of 2 inches or greater from safety-related equipment. Additionally, Section 4.1.4.3 directs Operations to determine if post-erection or removal inspections by Operations are warranted (i.e., scaffold location in close proximity to safety-related equipment).

Contrary to the above, from May 9 to May 22, 2023, Constellation personnel did not adequately accomplish scaffold construction, inspection, and approval in the A and C RHRSW room in accordance with approved scaffold control procedures. Specifically, scaffolding in this room was in direct contact with safety-related equipment. Additionally, corrective actions put in place to correct identical violations were not effective in ensuring that scaffolding erected to support maintenance functions would not impact safety-related equipment.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Provide Adequate Work Instructions in Safety-Related Breaker Maintenance Procedures Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.7] - 71152A Systems NCV 05000333/2023002-02 Documentation Open/Closed A Green self-revealed finding and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide adequate work instructions in the maintenance procedure used to service and inspect safety-related 4160 volt (V) breakers. Specifically, maintenance procedures MP-054.01 and MP-054.03, used for servicing the closing coil assembly in 4160 V breakers, did not provide adequate guidance to ensure proper fit and alignment of required fiber spacers. The incorrect installation of the spacers resulted in misalignment of the closing coil sleeve which caused the associated armature to bind against the spacer and the end of the sleeve, increasing the amount of force required to close the breaker, and resulting in the failure of breaker 71-10504 to close on demand on April 12, 2023. Additionally, breaker maintenance procedures do not include specific guidance to ensure breaker to cubicle alignment is performed and maintained, which could allow for breaker shifting to occur over time, and lead to interference issues between the breaker and the cubicle.

Description:

On April 12, 2023, EDG 'A' and 'C' tie breaker 71-10504 failed to close during planned surveillance testing. The failure of the breaker to close caused the 'A' EDG subsystem to be declared inoperable. The failed breaker was removed, a replacement breaker was installed, the surveillance test was performed successfully, and the 'A' EDG subsystem was returned to operable status.

The licensee initiated IR 04669409 to document the failure, and initiated a corrective action program evaluation (CAPE) in accordance with PI-AA-125, "Corrective Action Program Procedure," to investigate the cause of the failure. As part of the evaluation, the breaker was sent to the manufacturer for a failure analysis. The failure analysis concluded the following:

"The most probable cause of the circuit breaker failure to close on demand is the incorrect installation of a spacer within the closing coil assembly. The misaligned spacer did not allow the closing coil sleeve to seat properly in the cutout of the support. The combination of both issues decreased the clearance required for the closing coil armature to move freely inside the full length of the assembly. The decreased clearance at the end of the assembly caused the armature to bind against the spacer and the inside of the sleeve."

Following the failure analysis, Constellation determined that maintenance procedures used to inspect and service safety-related 4160 V breakers at FitzPatrick did not contain sufficient guidance to ensure that the closing coil assembly spacers were installed correctly, and that the fit and alignment were proper to ensure proper operation of the closing coil assembly.

Specifically, procedures MP-054.01 and MP-054.03 are used at FitzPatrick to perform inspections and service of 4160 V General Electric Magne-Blast breakers, and the guidance was not specific enough to ensure the proper installation of the spacers.

Extent-of-condition actions were initiated to inspect all other safety-related 4160 V breakers to ensure that the fiber spacers were installed and aligned properly to allow for interference free movement of the closing coil assembly. At the time of the inspection, no additional misaligned spacers had been identified on safety-related 4160 V breakers.

The failure of the tie breaker to close on demand impacted the ability of the 'A' and 'C' EDGs to force parallel. If the EDGs fail to force parallel, the tie breaker is tripped and the individual generator fields are flashed. The first generator to reach 90 percent rated voltage would close onto the bus, and the other generator would be blocked from auto closure.

In addition to the identification of the fiber spacer misalignment, it was also determined that the squareness and alignment of breakers and their associated cubicles are not checked and verified as part of maintenance or testing of the breakers. Alignment issues have been identified at FitzPatrick in the past following breaker failures as potential causes or contributing causes. For example, IR 04418483 documented an unexpected opening of breaker 71-10504 following an EDG 'A' subsystem maintenance outage. Investigation of that breaker failure determined that the likely cause was that the breaker was not "squarely racked in" and "the misalignment affected the aux contacts or housing limit switch for the breaker to function properly."

As part of the CAPE for the April 2023 failure, it was determined that Breaker to Cubicle Alignment should be proceduralized and performed to conform to preventive maintenance templates, the original equipment manufacturer's manual and Constellation fleet best practices. This failure to ensure proper procedural guidance is an additional example of the Constellation's failure to provide adequate work instructions in the maintenance procedure used to service and inspect safety-related 4160 V breakers.

Corrective Actions: At the time of failure, Constellation initiated troubleshooting of failed breaker 71-10504, replaced the failed breaker, and performed successful post-maintenance testing and surveillance testing. Subsequently, Constellation initiated actions to add guidance to breaker maintenance procedures, conduct extent-of-condition inspections on all safety-related 4160 V breakers, and schedule overhauls of all 4160 V safety-related breakers to be performed by the manufacturer.

Corrective Action References: IRs 04669409 and 04550060

Performance Assessment:

Performance Deficiency: Constellation failed to provide adequate instructions and acceptance criteria in procedures used to service and inspect safety-related 4160 V breakers.

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, Constellation did not provide adequate instructions in safety-related 4160V breaker maintenance procedures to ensure the closing coil assembly was properly assembled. This led to an adverse impact to the reliability of the breakers, and on April 12, 2023, EDG tie breaker 71-10504 did not close on demand during surveillance testing.

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 used Exhibit 2 and determined that the finding represented a loss of a probabilistic risk assessment (PRA) system and/or function as defined in the Plant Risk Information Book or Constellation's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Exhibit 2, Question 5 answered "Yes").

Specifically, the EDG force-paralleling function credited for certain design-basis events and required a detailed risk evaluation.

The senior reactor analyst (SRA) made the following assumptions and adjustments:

• The exposure time is 30 days based on surveillance test intervals with the EDG and related breakers and would not degrade in standby.

• Failure of bus 10500 was used as a surrogate for the failure of tie breaker 71-10504. This breaker is not directly modeled in Standardized Plant Analysis Risk. The primary impact of failing to close (or remain closed) during a postulated event would be the potential loss of one bus (10500 for 'A' and 'C' EDGs) if a forced parallel between the paired emergency diesel generators, and subsequent automatic lock-out of one emergency diesel generator, occurred. This response prevents an overloaded emergency diesel generator condition during a loss of coolant accident (LOCA)event. Based on review of the Updated Final Safety Analysis Report and licensee PRA notebook this fails the response to LOCA based on maximum bus loading. Other events are not impacted by this response one emergency diesel generator would succeed in carrying non-LOCA loads.

• Using LOCA event trees, the SRA failed one bus (10500, ACP-BAC-LP-10500 set to TRUE) and increased the failure probability of other 4 kilovolt breakers by an order of magnitude to represent the potential increase in hardware failure.

• For non-LOCA event tree, the SRA increased bus 10500 and other 4 kilovolt breaker failure probabilities one order of magnitude.

• No recovery actions were considered.

• Nominal test and maintenance conditions were modeled.

• FLEX equipment was also considered (using 2022 Pressurized Water Reactor Owners Group failure data, and only N equipment). Results were insensitive to FLEX.

Assuming a 30-day exposure, for LOCA events the conditional core damage probability (CCDP) was estimated at 4.3E-7/year, which results in a delta core damage probability (delta CDP) of (4.3E-7/year CCDP - 2.9E-7/year baseline)

• 30/365 = 1.2E-8/year. For non-LOCA events the CCDP was estimated at 5.8E-6/year which results in a delta core damage probability (delta CDP) of (5.8E-6/year CCDP - 4.9E-6/year baseline)

• 30/365 = 7.4E-8/year. This results in a total delta CDP of 8.6E-8/year. For LOCAs, medium break LOCA was the dominant event followed by large break LOCA. Medium LOCA dominant cutset included depressurizing during the MLOCA and failure of HPCI to run followed by failures of loads supported by safety-related busses 10500/10600. The dominate core damage sequences for the non-LOCA events are loss of condenser heat sink, loss of main feedwater. The dominant cutset was failure to depressurize and failures of HPCI and reactor core isolation cooling.

Since total delta CDP was less than 1E-7/year, no external events or large early release frequency impacts were assessed. Therefore, this finding is characterized as an issue of very low safety significance (Green).

Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate and up-to-date documentation. Specifically, maintenance procedures used in servicing safety-related 4160 V breakers did not contain adequate guidance for technicians to ensure proper fit and alignment of required fiber spacers in the closing coil assembly.

Enforcement:

Violation: Procedures MP-054.01 and MP-054.03 are used at FitzPatrick to inspect and service 4160 V safety-related breakers. The information and guidance contained in those procedures was not adequate to ensure that fiber spacers installed as part of the closing coil assembly were aligned and allowing for interference free operation of the assembly.

Appendix B to 10 CFR Part 50, Criterion V, "Instructions, Procedures, and Drawings,"

requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to this requirement, prior to April 12, 2023, procedures used at FitzPatrick to service and inspect safety-related 4160 V breakers did not include appropriate quantitative or qualitative acceptance criteria to ensure that the closing coil assembly fiber spacers were installed correctly, or that the breakers and breaker cubicles were properly aligned to prevent interference in breaker operation.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Incomplete Mode Switch Position Verification Causes Group 1 Containment Isolation Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.11] - 71153 NCV 05000333/2023002-03 Challenge the Open/Closed Unknown A self-revealed Green finding and associated NCV of TS 5.4.1(a), Procedures, was identified because the reactor shutdown procedure did not contain adequate verification steps for Mode Switch repositioning. As a result, when operators failed to verify all indications for the Mode Switch following repositioning on September 26, 2022, a reactor scram and Group 1 containment isolation occurred.

Description:

The reactor Mode Switch establishes the operating mode of the reactor. It has four positions: Shutdown, Refuel, Start/Hot Standby, and Run. The Mode Switch ensures that safety-related systems are correctly aligned to support the various modes of operation for a boiling water reactor. Due to the number of systems and components that require different configurations, the pistol grip locking handle assembly drives four banks of contacts with a total of 64 contacts available for logic circuit operation.

On September 26, 2022, FitzPatrick operators were performing a planned reactor shutdown to commence a refuel and maintenance outage. During power reduction, at 2:30 AM the operator repositioned the Mode Switch from Run to Start/Hot Standby. At 3:06 AM a Group 1 isolation signal was received, resulting in the closure of all MSIVs and a reactor scram.

During the post transient review, an operator noted that the computer indication for the Mode Switch position did not update after it was moved from Run to Start/Hot Standby. Procedure HU-AA-101, "Human Performance Tools and Verification Practices," Revision 10, Step 4.1.1, states that self-check shall be used for component identification and equipment manipulations. Step 4.1.1.4 states in part, the individual will verify that the actual response is the expected response and if any unexpected response is obtained, then take actions as previously determined to ensure the system/component is in a safe condition and notify supervision.

Following the scram, FitzPatrick staff performed a follow-up investigation and troubleshooting.

The station identified that the scram was initiated by the MSIV closure RPS trip when reactor pressure was approximately 848 psig. The MSIV closure RPS trip should be bypassed when the Mode Switch is in Start/Hot Standby. As a result, the investigation concluded that the apparent cause was the Mode Switch contacts did not fully make up in the Start/Hot Standby position.

Procedure OP-65, "Startup and Shutdown Procedure," Revision 135, only includes a single means to validate the Mode Switch contacts are made up. FitzPatrick staff determined OP-65 did not contain adequate verification steps for Mode Switch repositioning, nor did it contain any precautions, limitations, or notes covering the operation of the Mode Switch.

Corrective Actions: FitzPatrick staff revised OP-65 to include steps to verify system response when repositioning the Mode Switch and added detail on how and why to operate the Mode Switch. Verifications include confirmation of the annunciator for MSIV closure trip in bypass in alarm, computer point verification, plant computer screen verification, and relay verification.

Corrective Action References: 04524574

Performance Assessment:

Performance Deficiency: FitzPatrick staff failed to maintain an adequate reactor shutdown procedure, OP-65, "Startup and Shutdown Procedure".

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the reactor shutdown procedure did not contain adequate verification steps for Mode Switch repositioning. As a result, operators failed to verify all indications for the Mode Switch following repositioning and a scram occurred as reactor pressure was reduced below the MSIV closure RPS trip setpoint.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) in accordance with Exhibit 1, because it did not cause a reactor trip coincident with the loss of mitigation equipment relied upon to transition the plant from the onset of a reactor trip to a stable shutdown condition.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, a computer point did not indicate a change in Mode Switch position following repositioning.

Procedure HU-AA-101 requires verification following component manipulation.

Enforcement:

Violation: 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 (RG) 1.33, Appendix A, November 1972. Regulatory Guide 1.33, Appendix A, Section B.9 requires General Plant Operating Procedures, which include shutdown.

Procedure OP-65, "Startup and Shutdown Procedure," Revision 135 only included a single means to validate the Mode Switch contacts are made up. FitzPatrick staff determined OP-65 did not contain verification steps for Mode Switch repositioning, nor did it contain any precautions, limitations, or notes covering the operation of the Mode Switch.

Contrary to the requirements of RG 1.33, procedure OP-65 did not contain verification steps for Mode Switch repositioning. As a result, on September 26, 2022, operators failed to verify all indications for the Mode Switch following repositioning and a scram occurred as reactor pressure was reduced below the MSIV closure RPS trip setpoint.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Minor Violation 71152A The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," for Constellation's failure to ensure that conditions adverse to quality associated with 4160 V safety-related breakers were promptly identified and corrected.

Specifically, PI-AA-125, "Corrective Action Program (CAP) Procedure," step 4.1.2, states:

If at any time (e.g., during an investigation, review of a CA closure, review of a previous CR)a SCAQ or CAQ [condition adverse to quality] or any question of either current or past Operability/Reportability arises, then INITIATE an Issue Report (IR) in accordance with PI-AA-120.

Procedure PI-AA-120, "Issue Identification and Screening Process," defines a CAQ as "an all-inclusive term used in reference to any of the following: failures, malfunctions, deficiencies, defective items, and non-conformances."

As part of IR 04669409, a CAPE was performed, which included a failure analysis performed by General Electric of failed breaker 71-10504. In addition to the concluded cause of the failure, General Electric also noted numerous other issues that were not related to the breaker's failure to close, but were identified as part of the failure analysis, and included:

• The racking screw cover plate was bent

• Disconnect/test/connect indicator return spring bolt loose and missing lock nut

• Mechanism to wheel support cage bolt loose

• Charging link binding

• Spring discharge toggle foot is below specification

• Close latch adjustment bolt not consistently returning to contact the mechanism frame

• Close latch switch to paddle clearance excessive

• Close latch switch paddle ground down on side to obtain clearance from the mechanism frame

• Upper arc chute to support resistance high on phase A and C

• Upper arch chute coil resistance high on phase C

• SAL 352.1 new lubricating grease not being used

• SAL 352.1 Replacement zinc plated shims with brass

• SAL 311.1 AMH Breaker discharge roller

• SAL 328.1 Magne-blast breaker ML-13 mech trip armature travel

• SAL 358.1 Lubrication recommendations

• SAL 360.1 HDO/AMH Manual trip lever

• SAL 360.1 New style ML-13/13A control switches

• SAL 362.1 Trip latch to roller clearance Some of the identified issues were considered to be "more significant" by General Electric and were specifically highlighted in the failure report, including:

• Opening spring seated off-center in the opening spring assembly

• Multiple mechanism bushings loose and falling out of the assembly, which will cause excessive wear of major mechanism components during normal circuit breaker operations

• Closing latch roller not properly lubricated, allowing the roller to move freely without the typical resistance provided by the grease

• No clearance between the trip latch and the trip latch roller with the mechanism in the reset position

• Wiring harness rubbing on charge/discharge operating arm

• High resistance measured on multiple aux switch contacts

• Movable contact hinge assembly has brass spacers in place of the required silver plated spacers on all phases, which are "mandatory for maintaining the continuous current rating of the circuit breaker," according to the failure analysis Inspectors determined that, following communication of these issues, Constellation did not initiate an IR to document the conditions for further evaluation, as required by PI-AA-125 and PI-AA-120.

Screening: The inspectors determined the performance deficiency was minor. The performance deficiency did not adversely affect the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The failure analysis that was performed determined that the noted issues were not related to the failure of the breaker, but were considered deficiencies and, in accordance with PI-AA-125 and PI-AA-120, were required to be documented and evaluated.

Enforcement:

The requirements in 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," state 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." Contrary to the above, Constellation failed to document deficiencies identified in a manufacturer-provided failure analysis that was performed following the failure of breaker 71-10504.

Constellation initiated IR 04686903 to document the issues identified in the failure analysis and had previously initiated actions to have safety-related 4160 V breakers overhauled by the manufacturer to address the identified deficiencies. This failure to comply with 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

Minor Performance Deficiency 71152S Minor Performance Deficiency: The inspectors evaluated a sample of issues and events that occurred over the first and second quarters of 2023. 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 corrective action program.

The inspectors had identified a trend in the fourth quarter 2022 associated with equipment reliability. The station generated IR 04546406 to address common causes with the equipment failures. The inspectors reviewed this IR as part of a 71152A inspection and the results are documented in this inspection report. During the inspection period for this sample, the station had experienced additional failures of safety-related 4 kilovolt breakers, which include the A core spray pump failing to start due to high resistance on the secondary disconnects of the associated 4160 V breaker on January 18, 2023, and the A and C EDG tie breaker failing to close due to closing coil misalignment on April 12, 2023. Also, the station experienced a partial containment isolation signal due to two separate failures of the reactor protection system power supply voltage regulator on April 14, 2023 and May 25, 2023.

The inspectors also noted the station identified an emerging trend associated with Operations personnel performance as documented in IR 04557800. This includes IR 04538719, Operations initiating water spray curtain 5 in the stairwell of the east crescent; a fire brigade member failing to attach their air regulator to their facepiece prior to entering the simulated hazardous air environment; and IR 04555544, an operator inadvertently initiating fire protection foam in the main condenser which entered into the reactor coolant system. The results of the inspection associated with IR 04555544 are documented in Inspection Report 05000333/2023001.

Screening: The inspectors determined the performance deficiency was minor. Based on the overall results of the semiannual trend review, the inspectors determined that Constellation had generally identified adverse trends before they could become more significant safety problems. 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.

Observation: Review of IRs 4550060 and 4669409: Common Cause Analysis 71152A and Corrective Action Program Evaluation of Safety-Related 4160 V Breaker Failures IR 04550060 was written on January 24, 2023, and initiated actions to perform a common cause evaluation of several recent equipment reliability issues. Inspectors reviewed the evaluation and associated IRs to assess whether Constellation performed the evaluation in accordance with established procedures.

Issue report 04669409 was written on April 12, 2023, and documented the failure of breaker 71-10504, EDG 'A' and 'C' tie breaker, to close during surveillance testing. The licensee performed a CAPE, which included a failure analysis of the breaker performed by the manufacturer.

Constellation determined through the common cause analysis and the CAPE that some safety-related equipment had not been appropriately classified in accordance with procedures, and that the preventive maintenance strategies did not align with fleet best practices. In addition, Constellation determined that breaker and cubicle alignment issues may have contributed to recent equipment failures. In response, Constellation initiated actions to verify the categorization of safety-related equipment, align preventive maintenance strategies with fleet best practices, and to verify and ensure alignment of breakers and breaker cubicles when being placed into service.

Inspectors identified a self-revealed violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," when Constellation failed to provide adequate instructions in maintenance procedures used to service and inspect safety-related 4160 V breakers. This NCV is dispositioned in the Inspection Results section of this report.

Additionally, inspectors identified a minor violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," for Constellation's failure to document deficiencies identified in a manufacturer-provided failure analysis that was performed following the failure of breaker 71-10504. This minor violation is dispositioned within the Inspection Results section of this report.

Observation: Outage-Related Valve Deficiencies 71152A The inspectors reviewed the adequacy of FitzPatricks problem identification and resolution process following unexpected failures and adverse conditions of several safety-related valves just prior to and during the fall 2022 refueling outage. This review included the assessment of adverse conditions and subsequent corrective actions for the following:

• 02-2MOV-53A, Reactor Water Recirculation Pump A Discharge Isolation Valve, motor failed during testing, IR 04527980

• 13MOV15, Reactor Core Isolation Cooling (RCIC) Steam Supply Inboard Isolation Valve, packing leak during hydrostatic testing, IR 04528860

• 13MOV21, RCIC Discharge Inboard Isolation Valve, pressure seal leak, IR 04526744

• 29BPV-1, Main Turbine Bypass Valve 1, elevated internal cylinder leakage, IR 04525955; and steam side inspection findings, IR 04525964

• 29MOV-74, Main Steam Inboard Line Drain Isolation Valve, overthrust event during diagnostic testing, IR 04527898

• 34NRV-111A, A Feedwater Supply Non-Return Valve, leakage from valve cover, IR 04527981 The inspectors reviewed work orders, weak-link analyses, diagnostic test results, non-destructive examinations, design control documents, as well as other engineering and maintenance activities performed on the valves to restore operability, as applicable. The inspectors evaluated the adequacy and timeliness of these corrective actions and maintenance activities.

The inspectors determined that the actions to address the individual valve conditions were reasonable and appropriate, and included maintenance, repair and testing to ensure operability, and applicable extent-of-condition reviews to ensure adverse conditions were not impacting similar valves. The inspectors did not identify any performance deficiencies or violations.

Observation: Incomplete Mode Switch Position Verification Causes Group 1 71152A Containment Isolation and Reactor Scram The inspectors reviewed Constellations Post Transient Review (PTR), Failure Modes Causal Table (FMCT), and CAPE associated with the reactor scram in September 2022 documented in IR 04524574.

The PTR identified that the reactor scram was unexpected for the plant conditions and determined that the Mode Switch, in the Startup/Hot Standby position, should have bypassed the main steam line low pressure Group 1 isolation of the MSIVs and the reactor scram on the MSIVs' "not full open" signal. The PTR stated that action will be taken as recommended by the FMCT (IR 04524574) to address the cause of the scram prior to plant startup. The troubleshooting tested all of the contacts in each of the four Mode Switch positions:

Shutdown, Refuel, Startup/Hot Standby, and Run. The Mode Switch was tested twice. The first time the Mode Switch was moved slowly from one position to the next, and the second time the Mode Switch was moved in a normal deliberate motion. The troubleshooting did not identify any failures of contacts in the Mode Switch.

At the time of the event, the Mode Switch was being troubleshot due to operating experience at another plant. In the technician's video recording of the movement of the gears in the Mode Switch, Operations was requested to move the Mode Switch from Run to Startup/Hot Standby slowly to improve the chances of seeing the internal movement of the Mode Switch.

The CAPE determined that the most probable cause of the scram is that the Mode Switch was not positioned exactly in the Startup/Hot Standby position, which did not bypass the main steam line (MSL) low pressure Group 1 isolation of the MSIVs and the reactor scram on the MSIVs' "not full open" signal. The CAPE identified that the slow movement of the Mode Switch probably caused the Mode Switch to be in an intermediate position between Run and Startup/Hot Standby, and therefore the switches did not make up to bypass the isolation and reactor scram.

The PTR requires that a timeline be established for the reactor scram. The site used the plant computer alarm messages to determine the sequence of events. They identified that at 3:05:43, MSL 'B' came in and at 3:06:17, MSL 'A' came in, which completed the logic for the Group 1 isolation. Then, 17 seconds later at 3:06:34, the reactor scram occurred. The FMCT identified that the Group 1 isolation occurred 7 seconds after the scram, at 3:06:41.

The inspectors identified that the time between the Group 1 isolation and the reactor scram should have been on the order of less than 1 second instead of 17 seconds. Also, the sequence of events presented by Constellation was inconsistent with the expected system response and was not identified by the PTR. Assuming the Mode Switch did not bypass the Group 1 isolation and MSIV scram, the alarm message sequence should have been in the order where low MSL pressure caused a Group 1 isolation which closed the MSIVs, which caused the reactor scram. The inspectors on May 23, 2023 inquired whether the site had a sequence of event (SOE) log which the site committed to in response to GL 83-28. The site then identified that they did have a SOE log which had not been used to determine the timeline and sequence of events. The site then attempted to obtain the file. Although the file was saved on a server, the SOE log did not generate the expected report with all the predefined points. Having the SOE log could have provided information that would have clarified the timing of the events and the sequence of events. The site wrote IR 04681926 to determine why the SOE log did not record the expected data. The site informed the inspectors that the 17-second discrepancy in the time between the second MSL signal and the scram was due to calibration differences between the MSL trip unit and the plant computer alarm message.

The CAPE identified that functions other than the MSL bypass and MSIV bypass were affected by the Mode Switch not being in Startup/Hot Standby, however they did not evaluate these functions because it was not part of the scope (cause of the scram) of the CAPE. The inspectors identified that they should have considered these other functions as part of the extent-of-condition. The inspectors questioned whether the average power range monitor (APRM) startup setpoint was operable (required in startup per TS) during the 36 minutes between when the Mode Switch moved from Run to Startup/Hot Standby. If the APRM startup setpoint was not operable for all APRMs this would have been a loss of safety function. The Mode Switch controls the scram setpoint of the APRMs. For TS 3.3.1.1, APRM Neutron Flux High (Startup), the TS setpoint is < or = 15 percent. In Run, the APRM Neutron Flux High (Fixed) TS setpoint is < or = 120 percent. In response, Constellation performed an evaluation and identified that with the Mode Switch between Run and Startup, the contacts that close in Run to select the 120 percent setpoint would have opened, causing the APRM trip setpoint to be 15 percent. Constellation reasonably established the APRM Startup setpoint was operable.

The inspectors noted the operator statements identified that when the Mode Switch was moved from Run to Startup/Hot Standby, the plant computer indicated that the plant mode was Run. The EPIC plant computer point D-89 (Startup) did not change state. The operators stated, the R-time indication was stuck on Run, indicating they thought that the computer indication was erroneous. They did not look at other available indications to determine if the Mode Switch was in Run. One of the other indications that should have been in was annunciator alarm window 09-1-5-12, "MSIV CLOSURE TRIP IN BYPASS." This violation is documented in this report.

EXIT MEETINGS AND DEBRIEFS

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

• On July 26, 2023, the inspectors presented the integrated inspection results to Timothy Peter, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Procedures AOP-7 Loss of 345 KV Breaker Air 18

ARP-09-8-6-11 LHH-FITZ 115 KV Line 3 BKR 10022 Air Lo or AC Pwr Loss 2

WC-JF-107-1000 Seasonal Readiness T&RM for JAF 5

71111.04 Corrective Action 04668457

Documents

Resulting from

Inspection

Drawings FM-21A Flow Diagram Standby Liquid Control System 11 37

FM-22A Flow Diagram Reactor Core Isolation Cooling System 13 57

Procedures OP-17 Standby Liquid Control System 55

OP-19 Reactor Core Isolation Cooling System 19

OP-21 Emergency Service Water (ESW) 42

OP-22 Diesel Generator Emergency Power 72

OP-43A 125 VDC Power System 35

OP-46B 120 VAC Power System

OP-60 Diesel Generator Room Ventilation 10

OP-62 Pipe and Cable Tunnels Ventilation Systems 16

71111.05 Fire Plans PFP-OUT1 Security Secondary Alarm Station (SAS) Fire Area/Fire Zone 4

YARD

PFP-PWR01 East Cable Tunnel/Elev. 258' Fire Area/Zone II/CT-2 3

PFP-PWR02 West Cable Tunnel/Elev. 258' Fire Area/Zone IC/CT-1 5

PFP-PWR22 Standby Gas Filter Room/Elevation 272', Fire Area/Zone 6

XX/SG-1

PFP-PWR31 Emergency Diesel Generator Spaces - South, Elev. 272' Fire 5

Area 5/Fire Zones EG-1, EG-2, and EG-5

PFP-PWR33 Pump Rooms (Screenwell)/Elev. 255' Fire Area/Zone XII/SP- 2

1, XIII/SP-2, IB/FP-1, FP-3

PFP-PWR34 Screenwell House and Water Treatment Area/Elev. 225', 5

255' and 260' Fire Area/Zone 1B/SH-1

PFP-PWR35 Screenwell House and Water Treatment Area/Elev. 272' Fire

Area/Zone 1B/SH-1

PFP-PWR42 Turbine Building-North/Elevation 252' Fire Area/Fire Zone 4

Inspection Type Designation Description or Title Revision or

Procedure Date

IE/TB-1

Procedures OP-AA-201-003 Fire Drill Performance 21

71111.11Q Procedures OP-18 Reactor Protection System 48

ST-24J RCIC Flow Rate and Inservice Test (IST) 56

71111.12 Corrective Action 04463794

Documents 04481502

04538630

04541686

Corrective Action 04671919

Documents

Resulting from

Inspection

Procedures ER-AA-320-1001 Maintenance Rule 18-10 - Scoping 0

71111.13 Corrective Action 04418483

Documents

Miscellaneous James A. Section 8.6 Emergency AC Power System 27

FitzPatrick

Nuclear Power

Plant Final Safety

Analysis Report

Procedures AOP-76 High Conductivity in Condensate System 9

OP-AA-108-117 Protected Equipment Program 32

ST-9BA EDG A and C Full Load Test and ESW Pump Operability 25

Test

Work Orders 5148113-01

71111.15 Corrective Action 04552383

Documents 04555907

04555914

04559521

04670504

04672634

04672886

Engineering 638298 Excess Cables Found on Duct in East Crescent 02/13/2023

Inspection Type Designation Description or Title Revision or

Procedure Date

Changes

Procedures CR-JAF-1999-

01993

CR-JAF-1999-

2063

RAP-7.4.01 Control Rod Scram Time Evaluation 33

SP-01.06 Gaseous Effluent Sampling and Analysis 22

Work Orders 5302853

71111.18 Procedures MA-AA-716-025 Scaffold Installation and Removal Request Process 19

71111.24 Corrective Action 04549971

Documents

Procedures ISP-22-2 High Pressure Coolant Injection (HPCI) System Loop Low 11

Flow Bypass Instruments

OP-27 Recirculation System 91

ST-2AM RHR Loop B Quarterly Operability Test (IST) 46

ST-3PB Core Spray Loop B Quarterly Operability Test 35

ST-9AB EDG System B Fuel/Lube Oil Monthly Test 14

ST-9BA EDG A and C Full Load Test and ESW Pump Operability 26

Test

Work Orders 5131119

5147028

5163279

5163280

5319488

5330698

5337724

5341611

5353487

5354686

5361367

71114.06 Procedures EP-AA-111 Emergency Classification and Protective Action 23

Recommendations

EP-AA-112 Emergency Response Organization (ERO)/Emergency 22

Inspection Type Designation Description or Title Revision or

Procedure Date

Response Facility (ERF) Activation and Operation

EP-AA-112-100 Control Room Operations 19

71151 Corrective Action 04685209

Documents

Resulting from

Inspection

Procedures LS-AA-2090 Monthly Data Elements for NRC ROP Indicator - Reactor 5

Coolant System (RCS) Specific Activity

ST-40D Daily Surveillance and Channel Check 126

71152A Corrective Action 04550060

Documents 04669409

Corrective Action 04686903

Documents

Resulting from

Inspection

71153 Corrective Action 04548050

Documents 04680732

Procedures AOP-14 Earthquake 22

AOP-60 Loss of RPS Bus B Power 15

OP-AA-108-111- Severe Weather and Natural Disaster Guidelines 26

1001

25