IR 05000298/2023004

Integrated Inspection Report 05000298/2023004

ML24038A256
Person / Time
Site:	Cooper
Issue date:	02/12/2024
From:	Jeffrey Josey NRC/RGN-IV/DORS/PBC
To:	Dia K Nebraska Public Power District (NPPD)
Josey J
References
IR 2023004
Preceding documents:	ML23214A274 (3 August 2023)
Download: ML24038A256 (37)

Text

February 12, 2024

SUBJECT:

COOPER NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000298/2023004

Dear Khalil Dia:

On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Cooper Nuclear Station. On January 24, 2024, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

Five findings of very low safety significance (Green) are documented in this report. Five of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with section 2.3.2 of the Enforcement Policy.

A licensee-identified violation which was determined to be of very low safety significance is also documented in this report. 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 violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Cooper Nuclear Station.

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 IV; and the NRC Resident Inspector at Cooper Nuclear Station. 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, Jeffrey E. Josey, Chief Reactor Projects Branch C Division of Operating Reactor Safety Docket No. 05000298 License No. DPR-46

Enclosure:

Inspection Report 05000298/2023004 w/attachment 1: Position Paper

Inspection Report

Docket No. 05000298

License No. DPR-46

Report No. 05000298/2023004

Licensee: Nebraska Public Power District

Facility: Cooper Nuclear Station

Location: Brownville, NE

Inspection Dates: October 1, 2023, to December 31, 2023

Inspectors: B. Baca, Senior Health Physicist G. Birkemeier, Resident Inspector K. Chambliss, Senior Resident Inspector

Approved By: Jeffrey E. Josey, Chief Reactor Projects Branch C 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 Cooper Nuclear Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. A licensee-identified non-cited violation is documented in report section 71111.1

List of Findings and Violations

Failure to Evaluate Fuel Pool Cooling System for (a)(1) Status in Maintenance Rule Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.2] - 71111.12 Systems NCV 05000298/2023004-01 Evaluation Open/Closed The inspectors identified a Green finding and non-cited violation of 10 CFR 50.65(a)(1), for the licensees failure to monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. Specifically, the licensee failed to evaluate a maintenance preventable functional failure of the fuel pool cooling system against established goals following the failure of positioner valve FPC-AOV-AO18B.

Failure to Make Necessary Adjustments during Required 10 CFR 50.65(a)(3) Evaluation Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.8] - 71111.12 Systems NCV 05000298/2023004-02 Procedure Open/Closed Adherence The inspectors identified a Green finding and non-cited violation of 10 CFR 50.65(a)(3), when the license failed to make necessary adjustments to the maintenance rule program.

Specifically, the licensee failed to evaluate multiple unplanned emergency core cooling actuations during the 24-month evaluation period.

Failure to Identify and Correct HVAC Maintenance Access Hatch Failures Resulting in Secondary Containment Inoperability Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.5] - Work 71111.15 NCV 05000298/2023004-03 Management Open/Closed The inspectors reviewed a self-revealed Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality, such as defective material and equipment are corrected. Specifically, the licensee failed to correct repeated failures of heating, ventilation, and air conditioning maintenance access hatches resulting in secondary containment vacuum level exceeding Cooper Nuclear Station technical specifications 3.6.4.1 limit on October 22, 2023.

Failure to Follow Radiological Control Procedures with Two Examples Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.12] - Avoid 71124.01 Radiation Safety NCV 05000298/2023004-04 Complacency Open/Closed The inspectors reviewed a self-revealed Green finding and associated non-cited violation with two examples of radiation workers failure to follow radiological control procedures.

Specifically, in the first example, a radiation worker failed to obtain permission from radiation protection prior to entering an area greater than 7 feet in the overhead. In the second example, a radiation protection technician failed to validate dose rates utilizing the heater bay remote area radiation monitors prior to allowing valve technicians to access the area during a power ascension.

Failure to Appropriately Analyze Supplying Steam to the Reactor Feed Pumps Using the Main Steam Line Drains Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.1] - 71151 Systems NCV 05000298/2023004-05 Resources Open/Closed The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee's failure to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to verify or check the adequacy of the design by performing an analysis or test that demonstrated that supplying steam to the reactor feed pumps using the main steam line drains would: (a) not exceed the allowable cooldown rate specified in the facility's technical specifications; and (b) would provide sufficient steam flow to allow the reactor feed pump to operate.

Additional Tracking Items

None.

PLANT STATUS

Cooper Nuclear Station began the inspection period at 80 percent power due to flexible power operations. The plant returned to rated thermal power on October 1, 2023. On October 13, 2023, power was lowered to approximately 80 percent for flexible power operations.

The plant returned to rated thermal power on October 14, 2023. On November 8, 2023, power was lowered to approximately 80 percent for flexible power operations. The plant returned to rated thermal power on November 9, 2023. On November 17, 2023, power was lowered to approximately 65 percent for a planned rod pattern adjustment. The plant returned to rated thermal power on November 18, 2023. On December 23, 2023, power was lowered to approximately 80 percent for flexible power operations. The plant returned to rated thermal power on December 24, 2023. The unit remained at 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 Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures for the following systems:

• fire protection

• service water

• off-gas

• ultimate heat sink

• standby gas treatment

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (1 Sample)

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

(1) division 2 service water system on November 1, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 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) off-gas building, 903-foot 6-inch elevation, on October 19, 2023

(2) turbine building reactor feed pump area, 882-foot 6-inch elevation, on October 31, 2023

(3) reactor building,1001-foot elevation, on November 8, 2023

(4) reactor standby gas treatment room and reactor motor generator set area, 976-foot elevation, on November 27, 2023

71111.06 - Flood Protection Measures

Flooding Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated internal flooding mitigation protections in the reactor building northwest quadrant on October 3, 2023.

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) division 2 residual heat removal heat exchanger on October 31, 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 and evaluated licensed operator performance in the control room during down power for flex operations from September 30, 2023, to October 1, 2023.

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

(1) The inspectors observed and evaluated licensed operator requalification simulator scenario training on November 21, 2023.

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) review of periodic evaluation in accordance with 10 CFR 50.65(a)(3) on December 22, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (1 Sample)

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) planned Yellow online risk during 4160 V, bus 1F, under voltage relay testing on November 1, 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) failure to include vendor required torque values for safety-related motor starter cover screws on October 13, 2023

(2) degraded condition of emergency diesel generator fuel oil cross-connect isolation valves on October 17, 2023

(3) failure to enter required technical specification limiting condition for operation condition during planned and emergent maintenance on residual heat removal service water booster pump subsystems on December 21, 2023

(4) drywell unidentified leakage flow transmitter failure leading to unplanned technical specifications entry on December 26, 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) control rod drive solenoid hydraulic control unit solenoid valve design change on October 6, 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) (3 Samples)

(1) reactor building fans vortex damper post replacement testing on November 3, 2023

(2) control rod drive solenoid-operated valve 117/118 replacement during down power on November 18, 2023

(3) division 2 diesel jacket water repairs post-maintenance test on November 30, 2023

Surveillance Testing (IP Section 03.01) (2 Samples)

(1) emergency diesel generator division 1 fuel oil day tank level switch on October 12, 2023

(2) reactor protection system turbine stop valve limit switch surveillance on November 22, 2023

Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)

(1) elevated trend in drywell unidentified leakage rate on November 2,

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.

Instructions to Workers (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.

Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)

The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:

(1) workers exiting the radiologically controlled area and surveying (small article monitor)personal and work items

(2) radiation protection technicians responding to and surveying workers and items for contamination post monitor alarms at the radiologically controlled area

(3) radiation protection technician surveying workers exiting contamination areas in the reactor and radwaste buildings

Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)

The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:

(1) radiation protection surveys of technical specification (locked) high radiation area (LHRA) and high radiation area (HRA) boundaries in the chemistry storage tank room

(2) radiation protection HRA and LHRA weekly boundary surveillance

(3) HRA and contamination area briefings for scheduled operator rounds and surveillances in the reactor building under radiation work permit (RWP) 2023-051, task1

High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)

The inspectors evaluated licensee controls of the following HRAs and LHRAs:

(1) radwaste/multipurpose facility HRAs: radwaste basement - lab drain tank, condensate backwash tank room, chemistry waste tank room, waste surge tank; and 903-foot elevation - filter demineralizer valve room, north door in the drum handling room, and south radwaste hallway

(2) radwaste/multipurpose facility LHRAs: radwaste basement - spent resin tank room, waste sludge tank room, phase separator tank room, floor drain/waste collector tanks room; and auxiliary radwaste 903-foot elevation - high integrity container storage pit, steel liner area

(3) reactor building HRAs: 859-foot elevation - 'E' sump pump in the southeast quad, southeast torus access, and northwest torus access; 881-foot elevation - three top of torus access points, northwest and southwest quads, and control rod drive (CRD)pump in southeast quad; 903-foot elevation - residual heat removal heat exchangers

'A' and 'B' room access; 931-foot elevation - residual heat removal heat exchangers

'A' and 'B' room access, fuel pool cooling piping at the north wall; and 958-foot elevation - reactor water clean-up (RWCU) valve room, inner control rod drive rebuild room

(4) reactor building LHRAs: 903-foot elevation - steam tunnel, drywell personnel access, and CRD hatch; 931-foot elevation reactor - RWCU heat exchanger room, phase separator pump room, and RWCU 'A' and 'B' pump rooms; and 958-foot elevation -

fuel pool cooling heat exchanger room

(5) turbine building HRAs:

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)

(1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.

71124.04 - Occupational Dose Assessment

Source Term Characterization (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.

External Dosimetry (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.

Internal Dosimetry (IP Section 03.03) (3 Samples)

The inspectors evaluated the following internal dose assessments:

(1) 2021 internal dose assessments for three workers

(2) 2022 internal dose assessments for four workers

(3) 2023 internal dose assessments for one worker

Special Dosimetric Situations (IP Section 03.04) (3 Samples)

The inspectors evaluated the following special dosimetric situations:

(1) five declared pregnant worker packages

(2) radiation work permit 2022-545, task 2, reactor disassembly and assembly effective dose equivalent monitoring

(3) radiation work permit 2022-554, task 1, condensate storage tank diving multipack application

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03)===

(1) July 1, 2022, through June 30, 2023 - As part of the inspection, the inspectors provided an NRC position paper to the licensee documenting the basis for the NRCs position that the December 16, 2022, scram should have been reported as a scram with complications. The position paper is included as attachment 1 in this report.

OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) October 1, 2022, through June 30, 2023

PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1) October 1, 2022, through June 30, 2023

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) review of corrective actions regarding fuel defect on November 3, 2023

(2) evaluation of licensee's control of fire doors on November 30, 2023

(3) review of licensee response to operator watches standing issues leading to delayed action in inserting manual scram due to a failed open turbine generator bypass valve on December 26,

INSPECTION RESULTS

Failure to Evaluate Fuel Pool Cooling System for (a)(1) Status in Maintenance Rule Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [P.2] - 71111.12 Systems NCV 05000298/2023004-01 Evaluation Open/Closed The inspectors identified a Green finding and non-cited violation of 10 CFR 50.65(a)(1), for the licensees failure to monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. Specifically, the licensee failed to evaluate a maintenance preventable functional failure of the fuel pool cooling system against established goals following the failure of positioner valve FPC-AOV-AO18B.

Description:

On October 22, 2022, during refueling outage RE32, positioner valve FPC-AOV-AO18B failed open draining water from the skimmer surge tank which was providing net positive suction head for the operating fuel pool cooling pumps. The fuel pool cooling system was providing the only credited method of decay heat removal at the time of the event. Operators had to take manual action to injecting core spray, a source of emergency core cooling, to prevent the loss of the fuel pool cooling system and decay heat removal capability. This event should have been classified by the licensee as a maintenance preventable functional failure.

The inspectors reviewed the Cooper Nuclear Station maintenance rule program and noted that station Procedure 3-EN-DC-206, "Maintenance Rule (a)(1) Process," revision 3C3, step 5.2 requires the system engineer to evaluate plant level events - such as core spray injection - within up to 90 days from the event date for an (a)(1) evaluation. This same step requires properly classifying the event against the station maintenance rule monitoring and established goals. The system engineer is required to develop the (a)(1) evaluation with a recommendation for placing the maintenance rule function either in (a)(1) or (a)(2) status.

During the inspection period, the inspectors questioned the system engineer and members of the maintenance rule expert panel about the fuel pool cooling system's maintenance rule status and existence of any (a)(1) evaluations for the system. The station responded by stating the core spray injection event and valve failure has been previously mischaracterized and no (a)(1) evaluation was performed.

The inspectors determined that this event should have been evaluated by the licensee as a maintenance preventable functional failure.

The licensee subsequently completed an (a)(1) evaluation on December 8, 2023, 413 days after the event. The (a)(1) evaluation recommended placing the fuel pool cooling function in (a)(1) status due to exceeding the performance criteria for plant level events. On December 19, 2023, the maintenance rule expert panel voted to move the fuel pool cooling function to (a)(1) status.

Corrective Actions: On December 8, 2023, a 10 CFR 50.65(a)(1) evaluation was completed under condition report CR-CNS-2023-04800. On December 19, 2023, the Maintenance Rule Expert Panel determined that the Maintenance Rule Function FPC-F01 for the fuel pool cooling system was to be moved to 10 CFR 50.65(a)(1) status and performance goals and monitoring criteria were identified. The licensee had previously performed an extent of condition for the fuel pool cooling system to ensure all components were properly scoped into the Maintenance Rule program.

Corrective Action References: Condition Reports CR-CNS-2022-05293, CR-CNS-2023-04800, and CR-CNS-2023-04905

Performance Assessment:

Performance Deficiency: Title 10 CFR 50.65(a)(1) requires, in part, that licensees shall monitor the performance or condition of structures, systems, or components, against licensee-established goals. The licensees failure to evaluate a maintenance preventable functional failure in accordance with Procedure 3-EN-DC-206, Maintenance Rule (a)(1)

Process, revision 3C3, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, similar to example 8.g of IMC 0612, Appendix E, "Examples of Minor Issues," the inspectors determined that the significance was more than minor because the Mitigating Systems cornerstone objectives were adversely affected because, when the Maintenance Rule functional failure was considered, performance indicates that the SSC was not being effectively controlled through appropriate preventive maintenance and that the SSC was not moved to 10 CFR 50.65(a)(1).

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding affected the qualification of a mitigating SSC and did not affect its operability or PRA functionality.

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee did not effectively evaluate the manual use of core spray injection or the failure of positioner valve FPC-AOV-AO18B to adequately determine the cause and significance of the event and the fuel pool cooling system's ability to maintain its ability to meet its safety function.

Enforcement:

Violation: Title 10 CFR 50.65(a)(1) requires, in part, that each holder of an operating license for a nuclear power plant shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions.

Contrary to the above, from October 22, 2022, to December 8, 2023, the licensee failed to monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions.

Specifically, the licensee failed to evaluate the fuel pool cooling system for 10 CFR 50.65(a)(1) status when a maintenance preventable functional failure of the positioner valve FPC-AOV-AO18B resulted in an unplanned emergency core cooling actuation of core spray.

Specifically, the licensee failed to evaluate a maintenance preventable functional failure of the fuel pool cooling system against established goals following the failure of positioner valve FPC-AOV-AO18B.

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

Failure to Make Necessary Adjustments during Required 10 CFR 50.65(a)(3) Evaluation Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.8] - 71111.12 Systems NCV 05000298/2023004-02 Procedure Open/Closed Adherence The inspectors identified a Green finding and non-cited violation of 10 CFR 50.65(a)(3), when the license failed to make necessary adjustments to the maintenance rule program.

Specifically, the licensee failed to evaluate multiple unplanned emergency core cooling actuations during the 24-month evaluation period.

Description:

On October 25, 2023, Cooper Nuclear Station completed a 24-month evaluation of the maintenance rule program. The evaluation covered the period of September 1, 2021, to August 31, 2023. The report was approved by the Cooper Nuclear Station maintenance rule expert panel and submitted to the NRC on October 31, 2023.

This report was prepared per the requirements of 10 CFR 50.65(a)(3) using guidance from NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," revision 4F, and Cooper Nuclear Station Procedure 3-EN-DC-207, "Maintenance Rule Periodic Assessment," revision 3C3.

Station procedure 3-EN-DC-207, step 5.4.2.4, states, in part, SSC performance related to plant level criteria should be assessed to determine maintenance effectiveness including reviewing condition reports and plant level monitoring per Procedure 3-EN-DC-205, "Maintenance Rule Monitoring," revision 7C0.

Station procedure 3-EN-DC-205, step 5.9, requires the system engineer to monitor plant level events including external indicator data, condition reports, licensee event reports, and operating data.

Thus, station procedures require the licensee to review all unplanned safety system actuations when evaluating the maintenance rule program and functions for effective monitoring and equipment performance.

During the inspectors review of the 10 CFR 50.65(a)(3) evaluation the inspectors noted that the report stated there were no unplanned emergency core cooling actuations during the evaluation period. The inspectors recalled that on October 22, 2022, Cooper Nuclear Station operators manually initiated core spray to recover level for a skimmer surge tank to maintain net positive suction head for fuel pool cooling pumps which were being utilized for decay heat removal for the spent fuel pool during refueling operations. This was the result of a maintenance preventable functional failure of the fuel pool cooling system. Additionally, the inspectors recalled that on November 12, 2022, the high-pressure coolant injection (HPCI)system was inadvertently initiated during reactor startup resulting in an automatic reactor scram. An inadequate modification to the HPCI system resulted in the unplanned actuation of the safety system. The inspectors determined that neither of these unplanned emergency core cooling actuations were evaluated as part of the (a)(3) evaluation preventing the licensee from making adjustments as necessary to the maintenance rule program to ensure the systems could perform their safety function.

The inspectors noted that both licensee event reports submitted to the NRC for each of the plant level events stated these events were unplanned emergency core cooling actuations.

The NRC Problem Identification and Resolution Inspection Report 05000298/2023010 (ML23135A107) issued on May 17, 2023, documented the core spray injection was unplanned, and the fuel pool cooling component failure was a maintenance preventable functional failure. NRC Integrated Inspection Report 05000298/2022004 (ML23024A125)issued on January 30, 2023, documented the HPCI actuation was unplanned. Both of these issues were documented in the associated reports as non-cited violations that were documented in the licensee's corrective action program and were required to be reviewed by the licensee when developing the (a)(3) report. The failure to properly review and evaluate the condition reports, licensee event reports, and non-cited violations resulted in the (a)(3) to be inaccurate and inadequate.

The inspectors interviewed members of the maintenance rule expert panel and the inspectors determined the members had either misunderstood or mischaracterized the events resulting in the plant level events' omission from the (a)(3) report.

The licensee entered these issues into the corrective action program to document the need to evaluate the plant level events.

On December 19, 2023, the Cooper Nuclear Station maintenance rule expert panel voted to resubmit a revised (a)(3) report.

Corrective Actions: Cooper Nuclear Station documented the NRC questions regarding the content of the (a)(3) evaluation. Additionally, Cooper Nuclear Station plans to resubmit a revised (a)(3) evaluation to the NRC. Additionally, the station performed a lesson learned to capture the failure to properly identify and evaluate plant level events.

Corrective Action References: Condition Reports CR-CNS-2023-04733 and CR-CNS-2023-04800

Performance Assessment:

Performance Deficiency: Title 10 CFR 50.65(a)(3) requires the licensee to perform an evaluation of their maintenance rule program and make adjustments as necessary. The failure to make necessary adjustments during their 10 CFR 50.65(a)(3) evaluation after failing to evaluate multiple unplanned emergency core cooling actuations during the 24-month evaluation period was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors determined the Mitigating Systems cornerstone objectives were adversely affected because failure to evaluate the unplanned emergency core cooling actuations reduced the reliability of the equipment.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding affected the qualification of a mitigating SSC and did not affect its operability or PRA functionality.

Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, the licensee failed to follow station procedures requiring all plant level events to be reviewed including licensee event reports to determine proper monitoring of maintenance rule functions.

Enforcement:

Violation: Title 10 CFR 50.65(a)(3) requires, in part, that adjustments shall be made where necessary to ensure that the objective of preventing failures of structures, systems, and components through maintenance is appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance.

Contrary to the above, on October 25, 2023, the licensee failed to make adjustments where necessary to ensure that the objective of preventing failures of structures, systems, and components through maintenance is appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance. Specifically, the licensees (a)(3) evaluation failed to consider two unplanned emergency core cooling actuations resulting from maintenance preventable functional failures which resulted in the licensee failing to evaluate the fuel pool cooling and high-pressure coolant injection system's preventive maintenance strategies.

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

Failure to Identify and Correct HVAC Maintenance Access Hatch Failures Resulting in Secondary Containment Inoperability Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.5] - Work 71111.15 NCV 05000298/2023004-03 Management Open/Closed The inspectors reviewed a self-revealed Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality, such as defective material and equipment are corrected. Specifically, the licensee failed to correct repeated failures of heating, ventilation, and air conditioning maintenance access hatches resulting in secondary containment vacuum level exceeding Cooper Nuclear Station technical specifications 3.6.4.1 limit on October 22, 2023.

Description:

On October 22,2023, a heating, ventilation, and air conditioning (HVAC)maintenance access hatch associated with reactor building HVAC exhaust fan EF-R-1B discharge duct failed open. This caused a perturbation in secondary containment pressure.

Secondary containment pressure reached -0.13-inch water gauge exceeding the Cooper Nuclear Station technical specifications 3.6.4.1.1 limit of -0.25-inch water gauge. This condition made secondary containment inoperable for 1 minute and 18 seconds before the reactor building HVAC system recovered secondary containment.

Upon receiving the reactor building high pressure alarm, CNS operators discovered the maintenance access hatch, located above the reactor recirculation motor generator B exciter, open. Shortly afterwards, the hatch was closed, and secondary containment pressure was restored to its normal setpoint of -0.33-inch water gauge. CNS system engineers performed a complete walkdown of every duct hatch within the plant.

The inspectors reviewed the licensee's corrective action program for similar hatch failures.

Seven prior failures of hatches associated with various systems within the station either failing open or falling off were identified between June 20, 2005, to October 22, 2023. In particular, the June 20, 2005, failure, which was captured in the licensees corrective action program as condition report CR-CNS-2005-04483, the hatch associated with HVAC exhaust fan EF-R-1A failed open resulting secondary containment pressure reaching 0.05-inch water gauge. These prior failures were corrected using Spot Maintenance with no further corrective actions taken.

Corrective Actions: The licensee entered the condition into its corrective action program and secured the failed maintenance hatch. Additionally, an extent of condition was performed for all maintenance access hatches on site. The licensee is subsequently developing a modification to ensure the maintenance hatches will not fail open.

Corrective Action References: Condition Reports CR-CNS-2023-04582, CR-CNS-2023-04770, CR-CNS-2023-04775, and CR-CNS-2023-04584

Performance Assessment:

Performance Deficiency: Title 10 CFR Part 50, appendix B, criterion XVI requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. The licensees failure to take appropriate corrective actions following the identification of multiple failures of HVAC maintenance hatches was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to correct the issues with the maintenance access hatches resulted in secondary containment pressure exceeding technical specifications limits.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power.

The inspectors determined that the finding was of very low safety significance (Green)because the finding only represented a degradation of the radiological barrier function provided for the control room, auxiliary building (reactor building), spent fuel pool, SBGT system (BWR), or EGTS system (PWR ice condenser).

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the station did not develop adequate restoration of the maintenance access hatches following maintenance activities leading to an unplanned inoperability of secondary containment.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are corrected.

Contrary to the above, from June 20, 2005, to October 22, 2022, the licensee failed to correct a condition adverse to quality associated with repeated failures of HVAC maintenance access hatches. Specifically, the station performed spot maintenance on the failures of the HVAC hatches but failed to correct or evaluate these conditions adverse to quality resulting in an unplanned loss of secondary containment.

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

Licensee-Identified Non-Cited Violation 71111.15 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: Technical specification (TS) 5.4.1.a requires, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, the licensee did not properly pre-plan and perform maintenance that can affect the performance of safety-related equipment in accordance with procedures, documented instructions, or drawings appropriate to the circumstances. Specifically, on July 28, 2022, and August 30, 2023, the licensee failed to develop a work package for maintenance on the service water booster pumps that either included temporary lateral seismic restraints or declaring the service water booster pump system inoperable in accordance with station Procedure 7.2.14, "RHR SWBP Overhaul and Replacement," revisions 46 and 47.

The licensee referenced Cooper Nuclear Station Procedure 7.2.57.1, "Pipe Support Removal and Reinstallation," revision 13, when developing a work package for maintenance on the safety-related service water booster pump system. This procedure is a generic maintenance procedure which directs the removal, reinstallation, and inspection of pipe supports.

Procedure 7.2.14, steps 2.2. and 2.3, have a precaution that states, if service water booster pump C or D are either removed or their respective discharge flange is disconnected then temporary lateral seismic restraints must be installed or the entire subsystem must be declared inoperable per Cooper Nuclear Station TS 3.7.1, condition B. Licensee calculation NEDC 00-024, revision 1, determined that when service water booster pump C or D is disconnected, the entire subsystem is no longer seismically qualified.

During preventive maintenance on service water booster pump D in July 2022, and corrective maintenance on service water booster pump C in August 2023, Cooper Nuclear Station did not install temporary lateral seismic restraints nor did the licensee enter TS 3.7.1, condition B.

In both instances, the maintenance activities were completed in less than the 7-day allowed outage time.

Significance/Severity: Green. The inspectors assessed the significance of the finding using Appendix A, "The Significance Determination Process (SDP) for Findings At-Power." The finding was determined to be of very low safety significance (Green) because it:

(1) was not a design deficiency or qualification impacting operability or PRA functionality;

(2) did not represent a loss of system and/or function;

(3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time;

(4) did not represent a loss of the probability risk analysis function of two separate technical specification systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;

(5) did not represent a loss of probability risk analysis system and/or function for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and

(6) did not result in the loss of a high safety-significant, nontechnical specification train.

Corrective Action References: Condition Report CR-CNS-2023-04524 Failure to Follow Radiological Control Procedures with Two Examples.

Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.12] - Avoid 71124.01 Radiation Safety NCV 05000298/2023004-04 Complacency Open/Closed The inspectors reviewed a self-revealed Green finding and associated non-cited violation with two examples of radiation workers failure to follow radiological control procedures.

Specifically, in the first example, a radiation worker failed to obtain permission from radiation protection prior to entering an area greater than 7 feet in the overhead. In the second example, a radiation protection technician failed to validate dose rates utilizing the heater bay remote area radiation monitors prior to allowing valve technicians to access the area during a power ascension.

Description:

First, on October 5, 2022, a radiation worker received a self-reading dosimeter dose rate alarm when trying to access valve RHR-MO-274B via scaffolding on the drywell 901-foot elevation. The scaffolding was approximately 16 feet in the overhead. The worker failed to receive prior permission from RP to enter the overhead area. The self-reading dosimeter had a peak dose rate of 218.1 millirem per hour (mR/hr) and the dosimeter dose rate alarm set point was 200 mR/hr. Surveys (CNS-2210-017 and CNS-2210-0103) of the scaffolding access for the area identified the residual heat removal and shutdown cooling lines were 500 mR/hr on contact and 370 mR/hr at 30 centimeters. The worker left the area upon receiving the dose rate alarm and reported to RP.

Licensee Procedure 9.EN-RP-100, Radiation Worker Expectations, revision 20, step 3.3, stated workers were not permitted to enter overhead areas above 7 feet without prior permission from RP. Specifically, the worker failed to inform the RP technician briefing them prior to entering the radiologically controlled area that they may enter the overhead as part of their work scope nor did the worker discuss the work scope change once it was understood the worker needed to access valve RHR-MO-274B from scaffolding in the drywell overhead.

Second, on November 16, 2022, a radiation worker (valve team technician) received a self-reading dosimeter dose rate alarm upon entering the heater bay to check for air supply line leakage on valve MSPCV-68. The valve team was working under Specific Work Permit 2022-054, Turbine Building High Radiation Areas - Specific Work Permit, revision 2.

The RP technician briefed the valve team on the radiological conditions consistent with the RP technicians entry into the same area earlier in the shift. At the time the valve team entered the heater bay, the radiological conditions had changed due to the planned power ascension. Upon entering the heater bay, one of the valve technicians received a dose rate alarm of 113 mR/hr with a dose rate alarm setpoint of 110 mR/hr. Both valve team members left the area and informed RP. The RP technician briefing the workers failed to validate the heater bay dose rates utilizing the remote area radiation monitors prior to allowing the valve team to access the area during the power ascension. A survey of the heater bay by an RP technician was not immediately taken as the dose rates were continuing to rise following the power ascension. Following the dose rate alarm, the licensee reviewed the area radiation monitors closest to the workers entry location and identified a general area radiation reading of approximately 120 mR/hr. This dose rate corresponds with the maximum dose rate of 113 mR/hr seen on the valve technicians dosimeter.

Licensee Procedure 9.NISP-RP-2, Radiation and Contamination Surveys, revision 4, step 7.2 stated, in part, RP coverage included the use of remote area radiation monitors to verify the radiation levels in steam affected areas prior to entry. Specifically, the RP technician did not validate dose rates utilizing remote area radiation monitors in the heater bay prior to allowing valve technicians to access the area during power ascension.

Corrective Actions: The licensee initiated corrective actions which included a stand down meeting with radiation protection staff and maintenance to discuss the operating experience, performed human performance evaluations and gap analyses, and enhanced supplemental training material and valve locator descriptions.

Corrective Action References: The condition was entered into the corrective action program as condition reports CR-CNS-2022-04394 and CR-CNS-2022-06243, respectively.

Performance Assessment:

Performance Deficiency: Radiation workers failing to follow radiological control procedures was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to follow procedures led to workers entering high radiation areas to encounter dose rates higher than briefed with RP technicians and were greater than the setpoints on their self-reading dosimeters.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix C, Occupational Radiation Safety SDP. The finding was determined to be of very low safety significance (Green) because the finding was not:

(1) related to as low as is reasonably achievable (ALARA) planning;

(2) did not involve an overexposure;

(3) did not involve a substantial potential for overexposure; and

(4) the ability to assess dose was not compromised.

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, in the first example, the radiation workers were complacent since they knew where the valve location was though the valve was vaguely described in the work order. In addition, the radiation workers requested fall protection from a roving RP technician who assumed the workers had permission to enter the overhead. In the second example, the RP technician was confident the power ascension was on hold (power ascension was in progress post refueling) and failed to verify the power ascension status prior to allowing the workers into a steam affected area.

Enforcement:

Violation: Technical specification 5.4.1 required written procedures be established, implemented, and maintained covering the activities in Regulatory Guide (RG) 1.33, revision 2, appendix A, February 1978. Appendix A, section 7(e)(1) of RG 1.33 required procedures, in part, for access control to radiation areas.

Licensee Procedure 9.EN-RP-100, Radiation Worker Expectations, revision 20, step 3.3, stated workers were not permitted to enter overhead areas above 7 feet without prior permission from RP.

Licensee Procedure 9.NISP-RP-2, Radiation and Contamination Surveys, revision 4, step 7.2, stated RP coverage included the use of remote area radiation monitors to verify the radiation levels in steam affected areas prior to entry.

Contrary to the above, on October 5 and November 11, 2022, respectively, radiation workers failed to follow procedures, as evidenced in the following two examples, which required workers to

(1) obtain permission prior to entering areas above 7 feet in the overhead and

(2) use remote area radiation monitors to verify the radiation levels in steam affected areas prior to entry for RP coverage purposes.

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

Failure to Appropriately Analyze Supplying Steam to the Reactor Feed Pumps Using the Main Steam Line Drains Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.1] - 71151 Systems NCV 05000298/2023004-05 Resources Open/Closed The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee's failure to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to verify or check the adequacy of the design by performing an analysis or test that demonstrated that supplying steam to the reactor feed pumps using the main steam line drains would:

(a) not exceed the allowable cooldown rate specified in the facility's technical specifications; and

(b) would provide sufficient steam flow to allow the reactor feed pump to operate.

Description:

On December 16, 2022, Cooper Nuclear Station operators manually scrammed the reactor due to lowering reactor pressure caused by a main turbine generator bypass valve failing open. During the NRC's review of the event and subsequent discussions with the licensee regarding the recovery of the plant, the licensee proposed the use of the main steam line drains to provide steam to the turbine driven reactor feed pumps (RFPs) as a method to restore feedwater.

During the December 2022 manual scram, a Group 1 automatic isolation occurred which closed all inboard and outboard main steam isolation valves (MSIVs). This secured the flow of steam from the reactor to the common main steam header downstream of the MSIVs. The station's RFPs operate via a steam-driven turbine to provide feedwater to the reactor and maintain inventory. A Group 1 isolation would therefore secure motive steam required to sustain operation of the RFPs and challenge the licensee's ability to maintain inventory during scram recovery. During the licensee's post scram report and risk analysis discussions with the NRC, the licensee posited the steam motive force could be supplied from the reactor to the RFPs using the main steam line drains through two motor-operated, throttle valves:

MO-78, which throttles steam to the main steam header; and MO-79, which throttles steam directly into the main condenser.

This method of supplying steam to the RFPs via MO-78 was presented to the NRC via summary discussions entitled "Procedural Steps for Cooldown with Condenser" and "Feedwater and Main Condenser Use While Main Steam Isolation Valves are Closed." These summaries stated that station Procedure 2.2.56, "Main Steam System," revision 56, section 5 supports Procedure 2.1.5, "Reactor Scram," revision 78, by lining up the main steam line drain path to provide steam to the secondary system. Procedure 2.1.5, attachment 4, steps 1.4 through 1.4.4.4, provide direction for the use of the main steam line drains and the condenser for cooldown. These steps specifically direct throttling MO-78 and MO-79 to control steam flow downstream of the MSIVs. The licensee stated that after re-establishing steam flow via the main steam line drains and throttling via the two motor-operated valves, Procedure 2.2.28.1, "Feedwater System Operation," revision 102, sections 23 and 24, direct the restoration of feedwater using the RFPs.

During the NRC's review of the licensee's post-scram report and a summary of the position of MO-78 and MO-79 during the event, a panel operator noted that MO-79 had to be repositioned several times to adjust for RPV cooldown rates and directed RPV pressure control bands. Specifically, the panel operator estimated that MO-79 was stroked open for approximately 5 seconds to establish a RPV cooldown path and had to be throttled shut to maintain cooldown rate.

In response to these summaries and review of station procedures, the NRC requested an analysis or calculation that would demonstrate the use of steam via the main steam drain lines to the RFPs to establish feedwater flow using MO-78 would:

(a) not exceed the allowable cooldown rate specified in the facility's technical specifications 3.4.9, which specifies the station's allowed pressure and temperature limits, including a 100F/hr cooldown rate; and

(b) would provide sufficient steam flow to allow the reactor feed pump to operate.

On October 30, 2023, the NRC held a meeting with the licensee to discuss these concerns.

During this meeting, the licensee stated the station did not have an analysis nor any supporting calculations or documentation demonstrating the use of the main steam line drains to provide motive force, which is described in station procedures.

Corrective Actions: The licensee entered the condition into their corrective action program and is evaluating their procedures providing steam to the reactor feed pumps using the main steam line drains for the potential to exceed any technical specifications limits and viability of the procedures.

Corrective Action References: Condition Reports CR-CNS-2023-04686 and CR-CNS-2023-04714

Performance Assessment:

Performance Deficiency: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Additionally, the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, using alternate or simplified calculational methods, or by the performance of a suitable testing program.

Specifically, the failure to evaluate the effect of supplying steam to the reactor feed pumps using main steam line drains is a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to verify a post-scram response procedure did not violate any technical specifications limits nor verified the adequacy of the procedure.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. This finding was screened to Green because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC) and the SSC maintained its operability and PRA functionality.

Cross-Cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. This finding had a cross-cutting aspect of resources, as described in the human performance cross-cutting area because organizational leadership did not ensure that station procedures were adequately developed and analyzed to support reactor safety.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.

Additionally, the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, using alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, from December 16, 2022, to December 31, 2023, the licensee did not assure that applicable regulatory requirements and design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to verify or check the adequacy of the design by performing an analysis or test that demonstrated that supplying steam to the reactor feed pumps using the main steam line drains would:

(a) not exceed the allowable cooldown rate specified in the facilities technical specifications; and

(b) would provide sufficient steam flow to allow the reactor feed pump to operate.

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 November 7, 2023, the inspectors presented the occupational radiation safety inspection results to Bill Chapin, General Manager of Plant Operations, and other members of the licensee staff.

• On January 24, 2024, the inspectors presented the integrated inspection results to Khalil Dia, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Corrective Action CR-CNS-2023-05040

Documents

Procedures 0-CNS-FAP-WM-Seasonal Reliability 1

016

2.4ICE River Icing 0

5.1Weather Operation During Weather Watches and Warnings 22

5.2SW Service Water Casualties 34

Work Orders WO 5411485

71111.04 Corrective Action CR-CNS-2023-04632

Documents

Drawings 2006, Sheet 1 Flow Diagram Circulating, Screen Wash & Service Water 91

Systems

2006, Sheet 4 Flow Diagram Control Building Service Water System 65

Procedures 2.2.71 Service Water System 136

71111.05 Corrective Action CR-CNS-2018-06108, 2020-03448, 2021-05742, 2023-01333

Documents

Fire Plans CNS-FP-221 Reactor Building - MG Set Area, Elevation 976 - 0 6

CNS-FP-222 Reactor Building - Standby Gas Treatment Room, Elevation 4

976 - 0

CNS-FP-223 Reactor Building - Refuel Floor, Elevation 1001 - 0 5

CNS-FP-243 Cooper Nuclear Station Fire Protection Pre-Fire Plan 6

Turbine Building Reactor Feed Pump Area Elevation 882-6

CNS-FP-267 Off-Gas Building Elevation 903-6 6

Miscellaneous Cooper Nuclear Station Fire Hazards Analysis Matrix - Fire 06/20/2002

Area I Zone 5B

Cooper Nuclear Station Fire Hazards Analysis Matrix - Fire 02/28/2003

Area I Zone 6

Procedures 0-BARRIER Barrier Control Process 40

0-BARRIER Barrier Control Process, Attachment 2 - Fire Barriers (Fire 40

Protection Engineering)

0-BARRIER-Barrier Maps 13

MAPS

Inspection Type Designation Description or Title Revision or

Procedure Date

0-BARRIER-Miscellaneous Buildings 8

MISC

0.23 CNS Fire Protection Plan 92

0.7.1 Control of Combustibles 43

2.2.30 Fire Protection System 83

6.FP.302 Automatic Deluge and Pre-Action Systems Testing 36

6.FP.604 Fire Door Full Examination 35

7.3.21.10 Fire Door Preventative Maintenance Procedure 19

71111.06 Corrective Action CR-CNS-2023-04371

Documents

NDE Reports NEDC 09-102 Internal Flooding - HELB, MELB, and Feedwater Line Break3

71111.07A Corrective Action CR-CNS-2019-02541, 2019-02587, 2020-00333

Documents

Drawings 2040, Sheet 2 Flow Diagram - Residual Heat Removal System Loop B 20

Procedures 3-EN-DC-316 Heat Exchanger Performance and Condition Monitoring 12C1

3.34 Heat Exchanger Program Implementation 20

Work Orders WO 5040498, 5167519, 5210282, 5211010, 5398034

71111.11Q Corrective Action CR-CNS-2023-00043, 2023-03663

Documents

Miscellaneous EP Scenario 7, Lesson Number: SKL054-03-008 2

Post Scenario Critique, EP Scenario 7, Lesson Number: 11/01/2023

SKL05403008

Procedures 2.1.10 Station Power Changes 123

2.1.5 Reactor Scram 78

2.4PC Primary Containment Control 23

2.4SRV Stuck Open Relief Valve 19

5.1QUAKE Earthquake 16

EOP-1A RPV Control (1-3) 23

EOP-3A Primary Containment Control (1-3) 20

EOP-7A RPV Level (Failure-to-Scram) 23

EPIPEALHOT CNS EAL Wall Chart Hot 22

71111.12 Corrective Action CR-CNS-2023-04733, 2023-04800, 2023-04902, 2023-05221, 2023-

Documents 05252

Inspection Type Designation Description or Title Revision or

Procedure Date

Miscellaneous WT-2021-0021-Maintenance Rule Periodic Assessment for Assessment 10/25/2023

24 Period September 1, 2021 to August 31, 2023

Procedures 3-EN-DC-204 Maintenance Rule Scope and Basis 4C0

3-EN-DC-205 Maintenance Rule Monitoring 7C0

3-EN-DC-206 Maintenance Rule (a)(1) Process 3C3

3-EN-DC-207 Maintenance Rule Periodic Assessment 3CE

71111.13 Procedures 0-CNS-FAP-OM-Emergent Issue Response, Risk Classification, and 1

031 Oversight Determination

0-CNS-WM-104 On-Line Risk Assessment 16

0-PROTECT-Protected Equipment Program 63

EQP

2.0.2 Operation Logs and Reports 123

2.0.3 Conduct of Operations 106

6.1EE.302 4160V Bus 1F Undervoltage Relay and Relay Timer 49

Functional Test (DIV 1)

71111.15 Calculations NEDC 00-0024 RHR SW Booster Pump Discharge Piping Operability with 1

Booster Pump Removed (Piping Disconnected)

Corrective Action CR-CNS-2004-02411, 2004-03033, 2005-09158, 2007-06286, 2009-

Documents 06279, 2023-03830, 2023-04092, 2023-04150, 2023-04151,

23-04277, 2023-04298, 2023-04370, 2023-04524

Miscellaneous Re-configure DGDO-V-19 from Open to Closed, Engineering 0

Evaluation 08-026

USAR Change Request 2009-003 11/07/2012

Normal Position for Manual Valves DGDO-V-22 and DGDO- 0

V-23, Engineering Evaluation 04-053

NEDC 97-012 Emergency Diesel Generator Fuel On-Site Storage 3

Technical Specification Requirements

VM-0568 1151 Series Pressure Transmitters & 1151 Smart Pressure 8

Transmitters

Procedures 12.5 CNS QC Functions 48

2.3_C-1 Panel C Annunciator C-1 37

6.2HV.303 Division 2 Essential Control Building Ventilation 22

Temperature Switch Change Out and Functional Test

7.2.14 RHR SWBP Overhaul and Replacement 47

Inspection Type Designation Description or Title Revision or

Procedure Date

7.2.57.1 Pipe Support Removal and Installation 13

7.2.71.1 Bolting and Torque Implementation 3

Work Orders WO 5399057, 5433802, 5464700, 5512675

71111.18 Corrective Action CR-CNS-2022-02415, 2022-06878, 2023-01423, 2023-03100

Documents

Miscellaneous Design Equivalent DRC-SOV-117 & SO-118 XX-XX Solenoid Valve Design 0

Change Package Change

DEC-5454361

NEDC 87-069AX Load Study for RPS Power Panel 1A CCN 1C4

NEDC 87-069AY Load Study for RPS Power Panel 1B CCN 1C5

VM-0023 CRD Hydraulic Control Units 20

VM-2021 Automatic Valve Composite Manual 9

Procedures 10.9 Control Rod Scram Time Evaluation 73

71111.24 Corrective Action CR-CNS-2022-02968, 2023-00213, 2023-00301, 2023-00435, 2023-

Documents 00766, 2023-00905, 2023-00925, 2023-00956, 2023-01654,

23-01720, 2023-02409, 2023-02431, 2023-02524, 2023-

03168, 2023-04083, 2023-04445, 2023-04939

Drawings 2002, Sheet 1 Burns & Roe, Flow Diagram Main, Exhaust & Auxiliary

Steam Systems

Miscellaneous VM-0406 Centrifugal Fans 5

Procedures 0-CNS-OP-109 Drywell Leakage 7

10.9 Control Rod Scram Time Evaluation 74

2.2.20.1 System Operating Procedure, Diesel Generator Operations 77

2.2.61 Primary Coolant Leakage Detection System 35

6.1DG.301 Fuel Oil Day Tank Level Switches Functional Test and 23

Solenoid Valve IST Closure Test (DIV 1)

6.2DG.101 Surveillance Procedure, Diesel Generator 31 Day 92

Operability Test (IST) (DIV 2)

6.RPS.302 Main Turbine Stop Valve Closure and Steam Valve 65

Functional Test

Work Orders WO 5416161, 5462978, 5487208, 5504602, 5523013

71124.01 ALARA Plans ALARA Case Study - Drywell Head Issues 0

Corrective Action CR-CNS-2020-03596, 2020-04143, 2020-04392, 2020-04712, 2020-

Documents 04909, 2020-05066, 2020-06244, 2021-01539, 2021-02198,

Inspection Type Designation Description or Title Revision or

Procedure Date

21-03027, 2021-04001, 2021-04015, 2021-04891, 2021-

04892, 2022-00723, 2022-01904, 2022-01960, 2022-02438,

22-02461, 2022-05493, 2022-05820, 2022-05893, 2022-

05893, 2022-06069, 2022-06243, 2022-06287, 2022-06474,

22-06503, 2022-06802, 2022-06908, 2023-00097, 2023-

00104, 2023-00127, 2023-00269, 2023-00671, 2023-00782,

23-00822, 2023-00846, 2023-01456, 2023-01924, 2023-

01972, 2023-02074, 2023-02348, 2023-03419

Corrective Action CR-CNS-2023-04687, 2023-04705

Documents

Resulting from

Inspection

Miscellaneous 2020 Station Alpha Analysis: Dry Active Waste, Control Rod 03/30/2021

Drive Mechanism, and Reactor Cavity

22 Station Alpha Analysis: Dry Active Waste and Main 03/07/2023

Steam Isolation Valves

RE-32 Radiological Department Post Outage Report 0

26708005 Waste Stream Results Review: Dry Active Waste 10/29/2020

577786003 Waste Stream Results Review: Spent Bead Resin 02/03/2022

603489001 Waste Stream Results Review: 2022 Refueling Outage RE-12/05/2022

10CFR61 Dry Active Waste Smears

CNS RP-139A Reactor Daily/Weekly LHRA and HRA checks: doors locked, 06/13/2023

properly posted, barriers in place

CNS RP-139B Radwaste/MPF Daily/Weekly LHRA and HRA checks: doors 06/12/2023

locked, properly posted, barriers in place

CNS RP-139C Turbine Building Daily/Weekly HRA and LHRA checks: 08/24/2023

doors locked, proper posting, barriers in place

WT-2023-0068-Performance Gap Analysis: RE-23 HRA Control Events 0

001

Procedures 7.4.32 Work Over, Near, or In Reactor Vessel, Dryer/Separator 20

Storage Pool, or Spent Fuel Storage Pool

9.ALARA.5 ALARA Planning and Controls 28

9.EN-RP-100 Radiation Worker Expectations 20

9.EN-RP-122 Alpha Monitoring 6

Inspection Type Designation Description or Title Revision or

Procedure Date

9.NISP-RP-02 Radiation and Contamination Surveys 4

9.NISP-RP-03 Radiological Air Sampling 5

9.NISP-RP-04 Radiation Protection Posting and Labeling 4

9.NISP-RP-05 Access Control for Radiologically Controlled Areas 2

9.RADOP.10 Radioactive Sources Control and Accountability 23

Radiation Status Board Turbine Building 932-foot elevation 10/17/2023

Surveys Status Board Auxiliary Radwaste 903-foot elevation 10/12/2023

Status Board Radwaste Basement East 09/20/2023

Status Board Radwaste Basement West 09/20/2023

Status Board Reactor Building 958-foot elevation 08/23/2023

Status Board Reactor Building 1001-foot elevation 09/27/2023

Status Board Reactor Building, Drywell Personnel Airlock, 10/26/2023

903-foot elevation

Status Board Turbine Building, Steam Air Jet Ejectors, 882-12/22/2020

foot elevation

Status Board Turbine Building, Steam Air Jet Ejectors, 882-06/13/2020

foot elevation

CNS-2310-0025 Radwaste Building 903-foot elevation Room 3002 10/30/2023

Self-Assessments LO-2022-0107 Radiological Hazard Assessment and Exposure Controls 07/27/2023

and Occupational Dose Assessment

71124.04 Corrective Action CR-CNS-2022-04048, 2022-04100, 2022-04367, 2022-04394, 2022-

Documents 04785, 2022-04943, 2022-05174, 2022-05616, 2022-05791,

22-05796, 2022-05813, 2022-05820, 2023-00605, 2023-

00774, 2023-01164, 2023-02757, 2023-03090, 2023-03449,

23-03648, 2023-04274, 2023-05050

Miscellaneous 2022 Internal Dose Prospectus 0

23 Internal Dose Prospectus 0

Procedures 9.ALARA.9 Prenatal Monitoring 1

9.EN-RP-203 Dose Assessment 11

9.EN-RP-206 Dosimeter of Legal Record - Quality Assurance 7

9.EN-RP-208 Whole Body Counting and In-Vitro Bioassay 5

9.NISP-RP-06 Personnel Contamination Monitoring 0

71151 Calculations Offsite Dose Assessment Manual Public Dose Calculations: 10/04/2023

Inspection Type Designation Description or Title Revision or

Procedure Date

23 1st and 2nd Quarters

Corrective Action CR-CNS-2022-00530, 2022-04111, 2022-05301, 2022-05303, 2023-

Documents 02429, 2023-02908, 2023-04686, 2023-04713, 2023-04714,

23-04715

Miscellaneous Performance Indicator Documentation and Data Review

Package

Cooper Nuclear Station Annual Radioactive Effluent Release 04/27/2023

Report January 1 through December 31, 2022

Procedures 0-CNS-LI-114 Regulatory Performance Indicator Process 0

0-EN-LI-114 Regulatory Performance Indicator Process 0

71152A Corrective Action CR-CNS-2021-04902, 2021-04922, 2021-04983, 2021-04985, 2021-

Documents 05051, 2021-05218, 2021-05246, 2021-05258, 2022-00842,

22-00996, 2022-01000, 2022-02042, 2022-04596, 2022-

294, 2022-06882

Engineering EE 09-035 Evaluation of Fire Doors 3

Evaluations

Miscellaneous Causal Analysis 2021 Fuel Defect in Cell 14-39 10/12/2023

Design Equivalent GNF2.02 Fuel Bundle Enhancements and NSF Fuel 1

Change Package Channels

NEDC 10-080 NFPA 805 Chapter 3 Fundamental Fire Protection Program 5

and Design Element Review

NEDC 14-043 Fire Safety Analysis for the Entire Power Block 2

Training Scenario Kick - Simulator Introduction Cycle 02-14 Presentation 266

SKLO120601

NEI 99-02 contains the following explanation and definition regarding unplanned scrams with

complications (USwC):

Purpose of the indicator:

This indicator monitors that subset of unplanned automatic and manual scrams that

either require additional operator actions beyond that of the normal scram or involve the

unavailability of or inability to recover main feedwater. Such events or conditions have

the potential to present additional challenges to the plant operations staff and therefore,

may be more risk-significant than uncomplicated scrams.

Indicator Definition:

The USwC indicator is defined as the number of unplanned scrams while critical, both

manual and automatic, during the previous four quarters that require additional operator

actions or involve the unavailability of or inability to recover main feedwater as defined

by the applicable flowchart (Figure 2) during the scram response (see definition of scram

response in the Definitions of Terms section) and the associated flowchart questions.

Definition of Term:

Scram Response refers to the period of time that starts with the scram and concludes

when operators have completed the scram response procedures, and the plant has

achieved a stabilized condition in accordance with approved plant procedures and as

demonstrated by meeting the following criteria.

For a BWR:

• No emergency operating procedure (EOP) entry conditions exist related to

either the primary containment or the reactor.

• Reactor cool-down rates are less than 100 degrees F/hr.

• Reactor water level is being maintained within the range specified by plant

procedures.

BWR Figure 2 Clarifying Notes: Was main feedwater not available or not recoverable

using approved plant procedures during the scram response?

If operating prior to the scram, did main feedwater cease to operate and was it unable to

be restarted during the reactor scram response? The consideration for this question is

whether main feedwater could be used to feed the reactor vessel if necessary. The

qualifier of not recoverable using approved plant procedures will allow a licensee to

answer NO to this question if there is no physical equipment restraint to prevent the

operations staff from starting the necessary equipment, aligning the required systems, or

satisfying required logic circuitry using plant procedures approved for use that were in

place prior to the scram occurring.

The operations staff must be able to start and operate the required equipment using

normal alignments and approved emergency, normal and off-normal operating

procedures. Manual operation of controllers/equipment, even if normally automatic, is

allowed if addressed by procedure. Situations that require maintenance or repair

activities or non-proceduralized operating alignments will not satisfy this question.

Additionally, the restoration of main feedwater must be capable of being restored to

provide feedwater to the reactor vessel in a reasonable period of time. Operations

should be able to start a main feedwater pump and start feeding the reactor vessel with

the main feedwater system within about 30 minutes from the time it was recognized that

main feedwater was needed. During startup conditions where main feedwater was not

placed in service prior to the scram, this question would not be considered, and should

be skipped.

NEI 99-02 also includes six questions applicable to boiling water reactor (BWR) scrams. If any

of the questions are answered Yes then the scram is counted as complicated. The Branch

concurs with Cooper Nuclear Stations positions, except for the following question:

5. Was main feedwater not available or not recoverable using approved plant procedures

during the scram response?

The NRC believes the answer should be yes and the licensee believes the answer should be no.

The licensee provides the following justification for their position:

Following the Reactor SCRAM, RPV water level peaked at +83" on the Steam Nozzle Level

Indicator and did not come back below 54" until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the initiating event. There was

no need or desire to restore a RFP as level was above Level 8 and below +110" and within the

range specified in Procedure 2.1.5 (Reactor Scram). There were no equipment challenges to

prevent restoration of an RFP using approved plant procedures. After stabilizing RPV pressure

following the PCIS Group 1 Isolation, the isolation signal was able to be reset at approximately

00:12 on 12/17/2022. Likewise, opening MSL drains (MS-MOV-MO74 & MS-MOV-MO77) was

performed without issue. With MSL drains opened, Procedure 2.1.5 (Reactor Scram) provides

direction to throttle MS-MO-78 (path to MSL Header) or MS-MO-79 (path to Main Condenser).

With this line-up, Main Steam was available to supply motive force to the RFPT via the MSL

Header or to control RPV Pressure and level utilizing a path to the Main Condenser. In addition,

Procedure 2.2.80 (Turbine High Pressure Fluid System) provides guidance to secure the DEH

pumps to allow closure of the Main Turbine Bypass Valves to ensure sufficient steam is

available to the RFPT. Procedure 2.2.28.1 (Feedwater System Operation) provides guidance for

restoration of the RFP allowing restart within 30 minutes. Based on these factors, there were no

impediments to Feedwater availability and it was recoverable using normal alignments and

operating procedures, though not needed.

Branch Position:

The branchs position is this question should have been answered yes and this event should be

counted as an unplanned scram with complications. The licensee had an equipment failure that

prevented restoration of the reactor feedwater pumps for approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> during the

scram response period, and to correct this issue the licensee had to engage in troubleshooting,

and then used SRO direction (not proceduralized guidance) to close an isolation valve to shut

bypass valve 1 (BV1).

The licensee states there were no equipment challenges to prevent restoration of a reactor feed

pump (RFP). However, the failure of turbine BV1 precluded sufficient pressurization of the main

steam line header to allow supplying steam to a reactor feed pump turbine (RFPT).

The licensee states if BV1 had been required to be closed to support operation of the reactor

feed pumps for feedwater injection, the operations staff would have isolated the supply of the

DEH fluid to BV1 utilizing guidance in Procedure 2.2.80, Turbine High Pressure Fluid System.

The branch notes that although operations procedure 2.2.80 provides direction to remove the

DEH system from service by placing the DEH pumps in Pull-to-Lock, it does not direct

performance of these actions for the purpose of shutting bypass valves or to address bypass

valve failure. The scram procedure (2.1.5) does not direct operations to enter procedure 2.2.80

to shut a bypass valve and operators are not trained to restore main turbine bypass valves in

the closed position by utilizing the DEH pump Pull-to-lock switch. Successful closure of the

bypass valve by this method is dependent on operators deducing entrance into

Procedure 2.2.80 for the purpose of bypass valve closure, and that the isolation of DEH fluid

would successfully mitigate the failure mechanism of the bypass valve, which was not known at

the time of scram recovery. The actual actions taken by the licensee following the scram were

the use of station procedure 7.0.1.7 to perform troubleshooting activities to diagnose the cause

of BV1 failing open, and to develop actions to recover the use of the steam system. Then the

licensee used SRO direction (not proceduralized guidance) to close valve TGF-V-58, BV-1 HI

PRESS FLUID SHUTOFF, which secured DEH to the failed bypass valve causing BV1 to shut,

restoring use of the steam system. This took approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to accomplish.

1The licensee stated in their FAQ that feedwater injection was neither required nor desired in

support of the transition to Hot Shutdown and 2that injection by HPCI and/or RCIC did not occur,

nor was it required during the SCRAM response. 3During the licensees scram review the control

room supervisor stated that he ordered the reactor operator to Scram the reactor and after the

Group 1 isolation, set priorities to establish RPV level on RCIC. He held off on this action due to

RPV level too high and would have caused RCIC to trip. At 2:00 a.m. Narrow and Wide range

water level returned on scale and at 2:43 a.m. following resetting of the high-water level trips,

operators started RCIC for RPV level and pressure control. Procedure 2.1.5 identified the use of

RCIC as a method for pressure and level control. The licensees control room logs identified that

the licensee did transition to General Operating Procedure 2.1.4.2, Hot Shutdown. This

procedure identified the RFP, which was unavailable due to MSIVs remaining closed, as a

method of level control. The procedure also identified RCIC as a method of both level and

pressure control. Procedure 2.1.4.2 was entered 28 minutes after RFP high level trips were

reset and the RFP, HPCI, and RCIC pumps were available. RPV water levels were lowering

during this period of time. The branch contends that the requirement of the indicator is that the

licensee be able to demonstrate that feed could have been restored within 30 minutes from the

time it was recognized that feedwater was needed. Based on actual plant conditions during the

scram response period the licensee was not able to restore feed using either MSIVs or MSL

drains while BV1 was failed open.

The licensee stated that MS-MO-78 (supply to main steam header) and MS-MO-79 (MSL drain

path the main condenser) were aligned and able to provide motive force to the RFPT via the

main steam line header. The branch notes that the licensees ability to restore steam to the

feedwater pumps using MS-MO-78 while remaining in compliance with their administrative and

technical specifications cool down rate limits has not been analyzed, and it is probable that the

cooldown rate would be exceeded while operations was performing actions to maintain

cooldown rates within the administrative and technical specification limits during SCRAM

recovery (throttling MS-MO-79). During discussions following the event the licensee informed

FAQ response to Lines 20 and 21 and Lines 31 through 37

FAQ top of page 8 of 8.

River Bend Post Scram Review dated December 16, 2022.

the NRC that Scram procedure 2.1.5, step 1.4.4.4, directed the throttling of MS-MO-79 as

necessary to control RPV pressure. They went on to state that while the exact positioning used

when opening MS-MO-79 is not available, the operator estimated that the MOV was stroked

open a maximum of 5 seconds. After this opening, MS-MO-79 had to be throttled several times

to keep RPV pressure stable and included the occasional full closure of the valve. To the best of

the operators' recollection, MS-MO-79 was never fully open.

4The licensees assumed timeline for completion of procedure 2.2.56, step 5, for the

pressurization of the main steam line header was provided to the branch. The total time to open

the MSIVs was determined to be 33 minutes using the actual time for the actions following the

scram and assumes that with a pressurization rate of 100 psig/minute, which is the maximum

allowable rate, pressurization of the main steam line header and opening of the MSIVs would

take 8 minutes. This did not consider the low power history and cooldown rates experienced

during the event which required throttling of steam flow through the steam line drains.

Additionally, the licensee assumed that the procedure was complete up to the step that opens

MS-MO-78 equalizing pressure across the inboard MSIVs. The 33 minutes did not consider the

actions associated with troubleshooting actions required to shut BV1 or alignment and startup of

the RF

P. As stated above, the closure of BV1 required troubleshooting to be completed, which

took 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Had the DEH pump Pull-to-Lock method to close BV1 been utilized,

troubleshooting to enter procedure 2.2.80 would have been necessary, as this method was not

explicitly referenced in procedures 2.1.5, procedure 2.2.56, or procedurally identified as an

action associated with a failure of BV1. The licensee also stated that procedure 2.2.28.1 would

be used to restart the RF

P. Due to the delay in restoration, the licensee no longer met the

requirements of a Quick or Hot restart. A Quick restart requires a restart less than 5 minutes

and a Hot restart requires less than 30 minutes following a RFP trip. Both times were exceeded

which required the licensee to use the normal startup in procedure 2.2.28 instead of faster

Quick or Hot restarts, further challenging the ability to restore reactor feedwater within the

minutes restoration time specified by NEI 99-02 due to the startup procedure and required

5warmup and speed ramp times identified in the procedure.

Based on the information discussed above, the branches position is that the unavailability of

and inability to recover feedwater during scram response is a qualifier for an USwC per

guidance in NEI 99-02. The licensee did not have feedwater available as an RPV injection

source during scram recovery and would not have been able to recover feedwater using only

approved plant procedures. Recovery of feedwater required the closure of BV1 to allow

sufficient pressurization of the main steam header to supply the RFPT. Isolation of DEH fluid to

the bypass valve was only accomplished by SRO direction (non-proceduralized action) after

maintenance troubleshooting was invoked. Troubleshooting Actions taken to close BV1 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />

after the valves failure are beyond the scope of the licensees normal SCRAM procedures

(2.1.5 - Reactor SCRAM). For the approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> the valve was failed open, feedwater

was not an available source for RPV injection and was not recoverable using approved plant

procedures within the 30-minute restoration time.

The licensee has failed to provide evidence to support the capability of operations to restore

main feedwater to the reactor vessel within a reasonable period of time utilizing normal

MSIV Re-Opening Times provided to the NRC on February 28, 2023. Provided insight for the time

required to re-open MSIVs following closure. Time assumed actual performance of procedure 2.2.56,

section 5, Opening MSIVs With Reactor Pressurized. Operators did not perform the actions of step 5.6.7,

which states, Ensure Turbine bypass valves closed.

Procedure 2.2.28, Attachment 1, Warmup Rates and Hold Duration Determination

equipment alignments and approved emergency, normal, and off-normal operating procedures

during this plant transient. Therefore, the failure of the bypass valve and subsequent actions

taken to restore the bypass valve to the closed position results in a Yes response for

feedwater not being available or recoverable using approved plant procedures during the scram

response. Thus, the Branch maintains that this event is an Unplanned Scram with

Complications.

A-5