IR 05000317/2022002

Integrated Inspection Report 05000317/2022002 and 05000318/2022002

ML22220A038
Person / Time
Site:	Calvert Cliffs
Issue date:	08/08/2022
From:	Brice Bickett NRC/RGN-I/DORS
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
References
IR 2022002
Download: ML22220A038 (17)

Text

August 8, 2022

SUBJECT:

CALVERT CLIFFS NUCLEAR POWER PLANT - INTEGRATED INSPECTION REPORT 05000317/2022002 AND 05000318/2022002

Dear David Rhoades:

On June 30, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Calvert Cliffs Nuclear Power Plant. On July 14, 2022, the NRC inspectors discussed the results of this inspection with Patrick D. Navin, and other members of your staff. The results of this inspection are documented in the enclosed report.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

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

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Calvert Cliffs 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 Calvert Cliffs 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 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Digitally signed by Brice Brice A. A. Bickett Date: 2022.08.08 Bickett 10:30:59 -04'00'

Brice A. Bickett, Branch Chief Projects Branch 3 Division of Operating Reactor Safety Docket Nos. 05000317 and 05000318 License Nos. DPR-53 and DPR-69

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000317 and 05000318 License Numbers: DPR-53 and DPR-69 Report Numbers: 05000317/2022002 and 05000318/2022002 Enterprise Identifier: I-2022-002-0031 Licensee: Constellation Energy Generation, LLC Facility: Calvert Cliffs Nuclear Power Plant Location: Lusby, MD Inspection Dates: April 1, 2022 to June 30, 2022 Inspectors: R. Clagg, Senior Resident Inspector L. Dumont, Senior Reactor Inspector S. Obadina, Resident Inspector G. Walbert, Project Engineer D. Werkheiser, Senior Reactor Analyst Approved By: Brice A. Bickett, Branch Chief Projects Branch 3 Division of Operating Reactor Safety Enclosure

SUMMARY The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Calvert Cliffs 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 Failure to Perform An Adequate Foreign Material Close-Out Inspection Results in a Unit 2 Manual Reactor Trip Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71153 NCV 05000318/2022002-01 Complacency Open/Closed A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1, Procedures, was identified when Constellation failed to perform an adequate foreign material close-out inspection for the 12 spent fuel cooling pump breaker cubicle as required by site procedures. Specifically, on November 21, 2021, Constellation failed to verify that the 12 spent fuel cooling pump breaker cubicle was free of foreign material as required by MA-AA-716-008, Foreign Material Exclusion Program, Revision 16, which resulted in a Unit 2 manual reactor trip.

Additional Tracking Items Type Issue Number Title Report Section Status LER 05000318/2022-001-00 LER 2022-001-00 for Calvert 71153 Closed Cliffs Nuclear Power Plant, Unit 2, Automatic Reactor Trip Due to High Reactor Coolant System Pressure LER 05000318/2021-004-00 LER 2021-004-00, Calvert 71153 Closed Cliffs Nuclear Power Plant, Unit 2, Manual Reactor Trip Due to Lowering Steam Generator Levels

PLANT STATUS Unit 1 operated at or near rated thermal power for the entire inspection period.

Unit 2 operated at or near rated thermal power for the entire 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 hot temperatures for the Units 1 and 2 emergency diesel generators and intake structures on May 12, 2022 71111.04 - Equipment Alignment Partial Walkdown (IP Section 03.01) (4 Samples)

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

(1) Unit 1, 12 and 13 saltwater pumps and associated equipment with 11 saltwater pump out of service for maintenance, April 29, 2022 (2) Unit 1, 1B emergency diesel generator, 0C diesel generator, and associated equipment with 1A emergency diesel generator out of service for maintenance, May 5, 2022 (3) Unit 1, 11 saltwater air compressor and associated equipment with 12 saltwater air compressor out of service for maintenance, June 7, 2022 (4) Unit 2, 21 emergency core cooling system and associated components with 22 emergency core cooling system air cooler out of service for maintenance, June 14, 2022

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit 1 high-pressure safety injection system on June 28, 2022.

71111.05 - Fire Protection Fire Area Walkdown and Inspection (IP Section 03.01) (5 Samples)

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

(1) Units 1 and 2, cable chases and control room complex, fire areas 20-24, May 12, 2022 (2) Units 1 and 2, horizontal chases and 69' elevation electrical rooms, fire areas 35-38, May 12, 2022 (3) 0C (station blackout) diesel generator building, fire area EDG0C, May 17, 2022 (4) Unit 2, east and west electrical penetration rooms, fire areas 26-27, May 25, 2022 (5) Unit 1, east and west electrical penetration rooms, fire areas 32-33, June 1, 2022 Fire Brigade Drill Performance (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the on-site fire brigade training and performance during an unannounced fire drill at the machine shop on June 8, 2022.

71111.07A - Heat Exchanger/Sink Performance Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Unit 1, 12 component cooling heat exchanger maintenance and testing, June 22, 2022 71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the control room during Unit 1, 'B' train safety injection valve testing on June 8, 2022.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed licensee operator training involving a steam leak inside containment, a loss of auxiliary feedwater common suction, and a loss of offsite power in the simulator on June 15, 2022.

71111.12 - Maintenance Effectiveness Maintenance Effectiveness (IP Section 03.01) (1 Sample)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) Unit 1, AR 4445774, control element assembly dropped during power ascension, reviewed on May 2, 2022 Quality Control (IP Section 03.02) (1 Sample)

The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:

(1) Unit 1, WO C93780548, 11A service water heat exchanger relief valve replacement, May 24, 2022 71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management (IP Section 03.01) (7 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) Unit 1, elevated risk condition due to 11 saltwater pump out of service for replacement, May 10, 2022 (2) Unit 1, elevated risk condition due to 1A emergency diesel generator out of service for maintenance, May 12, 2022 (3) Unit 2, elevated risk condition due to 22 component cooling heat exchanger out of service for maintenance, May 24, 2022 (4) Unit 1, elevated risk condition due to 11A service water heat exchanger out of service for maintenance, May 27, 2022 (5) Unit 2, risk informed completion time implementation due to channel 'A' reactor protection system issues, May 31, 2022 (6) Unit 1, elevated risk condition due to 12 containment air cooler out of service for maintenance, June 13, 2022 (7) Unit 1, elevated risk condition due to 1B emergency diesel generator out of service for maintenance, June 29, 2022 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) Unit 1, AR 4493073, U-4000-11, 14 kilovolt transformer, upper left fan blade sheared preventing rotation, April 19, 2022

(2) Unit 2, AR 4493548, gap between wall and frame for emergency hatch #2 (high energy line break barrier), May 24, 2022 (3) Unit 1, AR 4501988, 12 containment air cooler local control hand switch failed to operate in the high position, May 27, 2022 (4) Unit 1, AR 4501259, dissolved gas analyzer for U-4000-13, 4 kilovolt transformer, in alarm status, May 31, 2022 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) Unit 2, ECP-17-000204, Ovation turbine control megawatt transducer circuit modification, June 29, 2022 Severe Accident Management Guidelines Update (IP Section 03.03) (1 Sample)

(1) The inspectors verified the site's severe accident management guidelines were updated in accordance with the pressurized water reactor generic severe accident technical guidelines and validated in accordance with Nuclear Energy Institute 14-01, Emergency Response Procedures and Guidelines for Beyond Design Basis Events and Severe Accidents, Revision 1, on June 7, 2022.

71111.19 - Post-Maintenance Testing Post-Maintenance Test (IP Section 03.01) (8 Samples)

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

(1) Unit 1, WO C93756168, 1CV5159A, 12A service water heat exchanger saltwater strainer flushing valve replacement and testing, April 15, 2022 (2) Unit 1, WO C93702979, 11 saltwater pump replacement, April 28, 2022 (3) Unit 1, WO C93751562, perform emergency diesel generator-13 on 1A1 and 1A2 emergency diesel generator engines, May 6, 2022 (4) Unit 1, WO C93783627, perform speed control knob torque check of 11 auxiliary feedwater turbine governor, May 10, 2022 (5) Unit 1, WO C93752186, replace paper drive assembly for 1RE5280, containment atmosphere particulate radiation monitoring system, May 12, 2022 (6) Unit 1, WO C93649176, 13 saltwater pump motor replacement, May 24, 2022 (7) Unit 2, WO C93670498, 22 emergency core cooling system air cooler maintenance and testing, June 15, 2022 (8) Unit 2, WO C93783966, 21 instrument air dryer maintenance and testing, June 30, 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) (2 Samples)

(1) Unit 2, STP-O-008A(SA)-2, "Test of 2A DG and 4kV Bus 21 Undervoltage," Revision 35, April 28, 2022 (2) Unit 2, STP-O-5A22-2, "22 Auxiliary Feedwater Pump Quarterly Surveillance Test,"

Revision 9, May 5, 2022 Inservice Testing (IP Section 03.01) (1 Sample)

(1) Unit 1, WO C93782745, STP-O-65C3-1, "14 CAC SRW Inlet, 1CV1592, Quarterly Operability Test," Revision 2, May 12, 2022 71114.06 - Drill Evaluation Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)

(1) The inspectors observed and evaluated the conduct of an Emergency Preparedness drill involving the failure of a containment sump motor operator valve with an earthquake which causes flooding from the 12 saltwater header rupture and service water pump room, and the loss of the 14 safety-related bus with the loss of cooling accident on hot leg resulting in a General Emergency declaration on May 10, 2022.

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) (2 Samples)

(1) Unit 1, April 1, 2021 through March 31, 2022 (2) Unit 2, April 1, 2021 through March 31, 2022 71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02) (1 Sample)

(1) The inspectors reviewed the licensees corrective action program for potential adverse 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) (2 Samples)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000318/2021-004-00, Manual Reactor Trip Due to Lowering Steam Generator Levels, (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML22018A025). The inspection conclusions associated with this LER are documented in this report under Inspection Results.

(2) LER 05000318/2022-001-00, Automatic Reactor Trip Due to High Reactor Coolant System Pressure (ADAMS Accession No. ML22063A703). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to be foreseen and corrected and, therefore, was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified.

INSPECTION RESULTS Observation: Semiannual Trend Review 71152S The inspectors identified an adverse trend associated with degraded reactor coolant system Tcold resistance temperature detectors for the Unit 1 and 2 reactor protection systems.

Constellation entered the following issues into their corrective action program in order to be appropriately addressed.

• On December 8, 2021, as documented in AR 4465414, following surveillance testing Constellation identified that the Tcold temperature element for channel 'A' of the reactor protection system on Unit 2 was found to have failed low. Further investigation identified that the temperature element was degraded. The temperature element was replaced during a forced outage.

• On April 26, 2022, as documented in AR 4495809, control room operators received an unexpected alarm and determined that the Tcold indication for reactor protection system channel 'A' was erratic. Constellation replaced the temperature transmitter and power supply, and restored channel 'A' of reactor protection system on Unit 2 to service after satisfactory testing.

• On May 23, 2022, as documented in AR 4501587, Constellation identified that the Unit 2 reactor protection system channel 'A' Tcold indication was rising unexpectedly.

Constellation performed troubleshooting and determined that the temperature element needed to be replaced or repaired. The reactor protection system is designed such that two temperature elements provide indication, and the highest temperature is selected. Constellation removed the degraded temperature element from service and has actions to repair or replace the temperature element during the next refueling outage. The second temperature element is in service and providing appropriate indication.

• On June 28, 2022, as documented in AR 4508003, Constellation identified that the Tcold indication for channel 'C' of the reactor protection system on Unit 1 was indicating higher than normal. Constellation performed troubleshooting and determined that the power supply was degraded. Constellation replaced the power supply and returned channel 'C' of reactor protection system to service after satisfactory testing. In AR 4465778, Constellation documented a potential cause for the degraded Tcold indications as loose or degraded connections at the resistance temperature detectors quick disconnect wiring harnesses and has sent these items offsite for evaluation.

The inspectors review noted Constellation has an action to document the final determination of the cause for the degraded resistance temperature detectors due in August of 2022. In addition, the inspectors noted degraded power supplies have caused some of the Tcold

indication issues. In AR 4508003, Constellation has an action to review the reactor protection system power supply failure history and revise the preventive maintenance strategy as appropriate. The inspectors determined these deficiencies demonstrate an adverse trend with regards to degraded Tcold resistance temperature detectors and Constellation generated ARs 4465414, 4465778, 44995809, 4501587, and 4508003 to address the deficiencies. The NRC inspectors did not identify any findings or violations of more than minor significance during this review.

Failure to Perform An Adequate Foreign Material Close-Out Inspection Results in a Unit 2 Manual Reactor Trip Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71153 NCV 05000318/2022002-01 Complacency Open/Closed A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1, Procedures, was identified when Constellation failed to perform an adequate foreign material close-out inspection for the 12 spent fuel cooling pump breaker cubicle as required by site procedures. Specifically, on November 21, 2021, Constellation failed to verify that the 12 spent fuel cooling pump breaker cubicle was free of foreign material as required by MA-AA-716-008, Foreign Material Exclusion Program, Revision 16, which resulted in a Unit 2 manual reactor trip.

Description: On November 21, 2021, Constellation was in the process of performing post-maintenance testing on the 12 spent fuel cooling pump and breaker. When starting the pump, an arc flash event occurred in the breaker. As a result, the upstream breaker (24A 480 volt bus feeder breaker) tripped open. This caused a loss of bus 2Y10, which is a non-vital 120 volt AC instrument bus. The loss of bus 2Y10 led to the repositioning of Unit 2 condensate pump mini-flow valve, condensate booster pump mini-flow valve, and moisture separator reheater drain tank high level dump valve, which created a secondary transient. This resulted in the 21 steam generator feedwater pump (SGFP) tripping on low-suction pressure. The 23 SGFP started due to the trip of the 21 SGFP. However, the loss of bus 2Y10 also caused a loss of the control element drive system and created an inability for the station to manually move control rods, which is needed when using the 23 SGFP following a loss of either 21 or 22 SGFPs. At 10:46 am, without the ability to manually insert control element assemblies to lower reactor power, Unit 2 was manually tripped from 100 percent power due to lowering steam generator levels. Constellation manually initiated auxiliary feedwater to maintain level in the steam generators. From this event, Constellation submitted LER 05000318/2021-004-00.

The inspectors reviewed Constellation's root cause investigation, AR 4462339, and noted that foreign material in the 12 spent fuel cooling pump breaker cubicle caused the arc flash event. The foreign material was identified as insulating breaker cubicle stab covers that are used to protect plant personnel from energized parts. The root cause investigation stated that the installation of breaker cubicle stab covers without a tracking mechanism, which were not removed from the breaker cubicle following maintenance, was the root cause. The inspectors reviewed the procedures associated with breaker cubicle maintenance and electrical safety.

Specifically, the inspectors reviewed FTE-53A, Westinghouse 480 Volt Load Center Cubicle Maintenance, Revision 0002, which provides a warning that the stationary main stabs remain energized with 480 volt AC. In addition, the inspectors reviewed SA-AA-129, Electrical Safety, Revision 11, which provides direction for installing insulating covers for energized

parts. Specifically, it states, in part, that the licensee should install protective shields when a breaker is removed from a compartment and energized bus bars are exposed. The inspector noted that while the procedures provided direction for installing insulating covers, the procedures did not provide a tracking mechanism or removal instructions for the insulating breaker cubicle stab covers.

The inspectors reviewed WO C93680525 for the breaker cubicle maintenance and noted that it included documentation for the foreign material close-out inspection and also identified the work as a foreign material exclusion zone 2. The inspectors reviewed the MA-AA-716-008-F-01, Work Package Device and Close-out Form, Revision 1, which requires that a final close-out inspection be performed, where the individual verifies that the system/component is free of foreign material prior to final system closure. This form was completed in the work package indicating that the foreign material close-out inspection had been performed. The inspectors also reviewed MA-AA-716-008, Revision 16, which provided a definition for foreign material exclusion zone 2, which is a zone that is established in situations where a final visual internal inspection is possible prior to system closure. The inspectors also reviewed step 14.1.6, of MA-AA-716-008, Revision 16, which states, in part, that the licensee shall verify that the system internals, components, and parts being installed are free of foreign materials prior to reassembly. The inspectors determined that a foreign material close-out inspection should have identified the insulating breaker cubicle stab covers. The inspectors concluded that the licensee performed an inadequate foreign material close-out inspection, which resulted in a Unit 2 manual reactor trip.

Corrective Actions: Constellation revised procedure FTE-53A to prohibit the use of stab covers during breaker cubicle inspections. Constellation removed all stab covers from the shop and verified that supply chain does not have any future orders to purchase additional stab covers.

Corrective Action References: AR 4462339 Performance Assessment:

Performance Deficiency: Constellation's failure to perform an adequate foreign material close-out inspection was a performance deficiency. WO C93680525 classified the work as foreign material exclusion zone 2, which requires the use of licensee procedure, MA-AA-716-008, Foreign Material Exclusion Program, Revision 16. Specifically, step 14.1.6, states, in part, that the lead worker shall verify that the system internals, components and parts, being installed are free of foreign materials prior to reassembly. Contrary to this, Constellation failed to verify that the 12 spent fuel cooling pump breaker cubicle was free of foreign material, specifically, the insulating breaker cubicle stab covers in the breaker cubicle. As a result, an arc flash event occurred in the 12 spent fuel cooling pump breaker which subsequently required a Unit 2 manual reactor trip.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to perform an adequate foreign material close-out inspection of the 12 spent fuel cooling pump breaker cubicle resulted in an arc flash event inside the breaker and subsequently required a Unit 2 manual reactor trip.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened this finding for safety significance and determined that a detailed risk evaluation was required. Specifically, the finding caused a reactor trip and loss of mitigating equipment relied upon to transition the plant from a reactor trip to a stable shutdown condition.

The performance deficiency caused the loss of the 480 volt load center 24A, motor control centers 125/204R, and 120 volt instrument bus 2BUS2Y10. This resulted in a low-suction pressure trip of the 21 SGFP due to the loss-of-power failed-open position of minimum flow bypass valves and loss of manual reactor rod control (both due to loss of 2BUS2Y10),

necessitating a manual reactor trip by the operators. The subsequent loss of the 22 SGFP on high-discharge pressure after the reactor trip was not considered related to the performance deficiency. This initiating event was evaluated as a transient/reactor trip. However, the senior reactor analysts also evaluated the event as a loss of main feedwater to ensure a bounding sensitivity review was performed.

The regional senior reactor analysts used the Systems Analysis Programs for Hands-On Evaluation, Revision 8.2.6, and the Standardized Plant Analysis Risk Model for Calvert Cliffs Unit 2, Model Version 8.64, to conduct an initiating event analysis. This included consideration of the effect on systems, structures, and components that were impacted by the performance deficiency.

In accordance with IMC 0308 and Risk Assessment of Operational Events Handbook guidance, the Significance Determination Process evaluates the risk increase/significance of the performance deficiency that causes an initiating event by using the incremental conditional core damage probability estimate:

Incremental conditional core damage probability = Conditional core damage probability -

Baseline core damage probability The initiating event was set to 1.0, including any inclusive systems, structures, and component failures, and the conditional core damage probability was calculated by the Systems Analysis Programs for Hands-On Evaluation. For this finding, the initiating event transient/reactor trip was set to 1.0. For the loss of main feedwater sensitivity case, event loss of main feedwater was set to 1.0. The following additional standardized plant analysis risk model modifications were made:

• 21 SGFP (MFW-TDP-FR-PUMP21) set to 0.2, as a surrogate for recovery, (21 SGFP was recovered shortly after instrument bus 2Y10 was cross-tied to 2Y09, which provided power to the min-flow valves restoring suction pressure)

• AC Bus 24A (ACP-BAC-LP-24A) set to TRUE, (bus was restored 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after event). This also fails motor control center 125 / 204R and associated loads

• Instrument Bus 2Y10 (ACP-BAC-LP-2Y10) failure probability was increased by an order of magnitude, as a surrogate for recovery, (2Y10 was cross-tied to 2Y09 in 7 minutes per the direction of abnormal operating procedure (AOP-07I-2))

• After discussions with Idaho National Labs, motor control center 204R fault tree was modified to account for short term recovery probability to cross-tie to motor control center 214R via procedure. A Standardized Plant Analysis Risk-H human error probability was calculated to be 0.14 for high-stress, moderate complexity, and

minimum time to complete. (conditional core damage probability results were insensitive to variations to this basic event)

• (Sensitivity) Alignments of key equipment were set to run TRUE or standby FALSE.

This was performed to realistically assess the important risk contributors due to the cross-train capability and flexibilities modeled in the Calvert Cliff Unit 2 standardized plant analysis risk model The following influential assumptions were used:

• Transient/reactor trip best represented the initiating event

• 21 SGFP loss was directly related to the performance deficiency; 22 SGFP was lost after the reactor trip due to an unexpected feedwater system response after receiving its expected reactor trip override and speed runback and is not considered related to this performance deficiency

• Standby electric feedwater pump (23 SGFP), which started once the 21 SGFP was lost, subsequently tripped as designed, coincident with the reactor trip and hence is considered to have little to no mitigating impact to the initiating event

• Recovery of feedwater and buses 2Y10/125/204R were feasible and justified

• Bus 24A was not readily recoverable based on troubleshooting and repairs. The licensee provided a damage assessment and justification for possible early recovery, but the senior reactor analysts did not consider this in the risk assessment

• Nominal test and maintenance values; A sensitively case was performed with key equipment alignments set as existed during the event (loss of main feedwater, service water, and switchgear ventilation)

Senior reactor analysts preliminary risk assessments determined the finding to be risk significant and identified dominant event trees as medium-break loss of coolant accidents and anticipated transients without scram. The dominant sequences involved loss of reactor coolant pump seals, failure to trip reactor coolant pumps, and high-pressure injection failure.

This was similar for both transient/reactor trip and loss of main feedwater (sensitivity) initiating event assessments. Further review of dominant sequence cut sets indicated this was mainly the result of motor control center 204R failure, loss of component coolant water, and emergency switchgear cooling that results in failure of emergency switchgear loads. This makes the conditional core damage probability results highly sensitive to alternate switchgear ventilation (as proceduralized in OI-22H) and the failure probability of the portable generators powering those fans. A review of calculation CA04570, that documents heat-up of the Unit 1 and 2 switchgear rooms on loss of ventilation, indicates that up to 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> is available to establish alternate ventilation before exceeding operational temperature limits for the affected emergency switchgear. Therefore, the senior reactor analysts concluded that this supported a low failure probability of the emergency switchgear loads from a loss of ventilation. Based on additional discussions with Constellation and considering the last bus was recovered within 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the event, additional standardized slant analysis risk model adjustments were made:

• Alternate emergency switchgear ventilation power (ACP-MGN-FC-PORTGEN) was reduced by two orders of magnitude to account for the large number of units tested and available

• Operator action to establish alternate emergency ventilation (HVC-XHE-XM-RMCOOL) was reduced by an order of magnitude to account for the large available time to establish the alternate line-up

• Various basic event probability updates based on senior reactor analyst's discussions with licensee probabilistic risk assessment staff to account for plant recoveries over the 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the event The calculated conditional core damage probability for a transient with the above adjustments was 8.2E-7. The baseline core damage probability was calculated to be 2.1E-7.

The incremental conditional core damage probability was 8.2E-7 - 2.1E-7, which resulted in a risk increase 6.1E-7. The dominant sequence is a transient, with loss of reactor coolant pump seals (from the loss of component cooling water), failure to trip reactor coolant pumps, and failure of high-pressure injection.

A sensitivity case treating the event as a loss of main feedwater was conducted and resulted in similar conditional core damage probability results and dominant sequences. The loss of main feedwater case was sensitive to the 21 SGFP recovery and availability of low-pressure feedwater. Both transient/reactor trip and loss of main feedwater risk analyses were insensitive to crediting diverse and flexible mitigation strategies (i.e., FLEX).

The performance deficiency did cause a transient (reactor trip), however an external event that could cause a transient cannot cause the performance deficiency to be revealed. The senior reactor analysts validated through discussions with NRC inspectors, review of the licensees root cause analysis, etc. that the performance deficiency will only have a risk impact when connecting the affected breaker to the bus. Therefore, the risk can be assessed looking only at internal events and risk contribution due to fire, flood, tornado, or seismic events need not be considered.

A large early release frequency assessment was made using the Systems Analysis Programs for Hands-On Evaluation and IMC 0609, Appendix H, dated March 23, 2020. The Systems Analysis Programs for Hands-On Evaluation calculated increase in large early release frequency was <1E-7 based on zero-factor multipliers for dominant sequences. Also, since Calvert Cliffs Unit 2 is a combustion engineering designed pressurized water reactor a separate consequential steam generator tube rupture screening was performed in accordance with IMC 0609, Appendix H, Section 5, aided by NUREG-2195, Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants, Appendix L, dated July 2017. Since all dominant sequence large early release frequency factors were zero, this assessment screened out having very low safety significance Constellation also performed an analysis of the performance deficiency using the Calvert Cliffs Unit 2 full-power internal events application-specific model based on the model of record and provided it to the analysts for information. The senior reactor analysts reviewed the analysis, key inputs, and assumptions and determined that their analysis was consistent with the NRC method. The licensees calculated incremental conditional core damage probability was 7.83E-7 with similar dominant sequences identified by the senior reactor analysts evaluation.

In summary, the increase in risk associated with the performance deficiency using the incremental conditional core damage probability estimate is 6.1E-7 and represents a finding of very low safety significance (Green). Incremental conditional core damage probability or delta core damage frequency are the risk metrics for the Significance Determination Process to evaluate the significance of inspection findings, and their numerical values are consistent with the risk informed scale and basis detailed in IMC 0308.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, Constellation failed to recognize the task of installing insulating breaker cubicle stab covers introduced foreign material into the breaker cubicle and should have recognized and planned for the possibility of mistakes.

Enforcement:

Violation: The Renewed Facility Operating License for Calvert Cliffs Power Plant, Unit 2, Technical Specification 5.4.1 requires, in part, that written procedures shall be established, implemented, and maintained as covered in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, where Section 9a specifies 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. MA-AA-716-008, Foreign Material Exclusion Program, Revision 16, step 14.1.6 states, in part, that the lead worker shall verify that the system internals, components, and parts, being installed are free of foreign materials prior to reassembly.

Contrary to the above, on November 21, 2021, Constellation failed to satisfactorily implement MA-AA-716-008 for the 12 spent fuel cooling pump and breaker maintenance. Specifically, Constellation did not perform an adequate foreign material close-out inspection and left the insulating breaker cubicle stab covers installed in the breaker cubicle. This resulted in an arc flash event inside the breaker and subsequently required a Unit 2 manual reactor trip.

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

EXIT MEETINGS AND DEBRIEFS The inspectors verified no proprietary information was retained or documented in this report.

• On July 14, 2022, the inspectors presented the integrated inspection results to Patrick D.

Navin, and other members of the licensee staff.

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