IR 05000461/2025010

Comprehensive Engineering Team Inspection Report 05000461/2025010

ML25128A110
Person / Time
Site:	Clinton
Issue date:	05/08/2025
From:	Karla Stoedter NRC/RGN-III/DORS/EB2
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
References
IR 2025010
Download: ML25128A110 (1)

Text

SUBJECT:

CLINTON POWER STATION - COMPREHENSIVE ENGINEERING TEAM INSPECTION REPORT 05000461/2025010

Dear David Rhoades:

On May 1, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Clinton Power Station and discussed the results of this inspection with D. Shelton and other members of your staff. The results of this inspection are documented in the enclosed report.

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

A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as 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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at Clinton Power 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 III; and the NRC Resident Inspector at Clinton Power Station.

May 8, 2025 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, Karla K. Stoedter, Technical Support Team Leader Technical Support Staff Division of Operating Reactor Safety Docket No. 05000461 License No. NPF-62

Enclosure:

As stated

Inspection Report

Docket Number:

05000461

License Number:

NPF-62

Report Number:

05000461/2025010

Enterprise Identifier:

I-2025-010-0021

Licensee:

Constellation Nuclear

Facility:

Clinton Power Station

Location:

Clinton, IL

Inspection Dates:

January 27, 2025 to May 01, 2025

Inspectors:

M. Abuhamdan, Reactor Inspector

J. Corujo-Sandin, Senior Reactor Inspector

S. Gardner, Contractor

T. Hartman, Senior Project Engineer

E. Magnuson, Reactor Inspector

L. Rodriguez, Senior Reactor Analyst

E. Rosario, Trans & Storage Safety Inspector

Approved By:

Karla K. Stoedter, Technical Support Team Leader

Technical Support Staff

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Comprehensive Engineering Team Inspection at Clinton Power 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: 71111.21M.

List of Findings and Violations

Failure to Perform Walkdowns Required by ASME Code Case N-789 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-01 Open/Closed

[H.5] - Work Management 71111.21M The inspectors identified two examples of a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR 50.55a(b)(5) for the failure to comply with the requirements of ASME Code Case N-789-3 and N-789-5. Specifically, the licensee failed to monitor pressure pads installed on ASME Code Class 3 piping for leakage every 31 days in accordance with ASME Code Case N-789.

Failure to Perform Appropriate Testing of Main Steam Isolation Valve 1B21-F022D Reactor Protection System Instrumentation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-02 Open/Closed

[H.3] - Change Management 71111.21M The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure testing required to demonstrate the 1B21-F022D main steam isolation valve reactor protection system instrument would perform satisfactorily in service was identified and performed in accordance with written test procedures, which incorporated the requirements and acceptance limits contained in applicable design documents and for the failure to evaluate the test results to assure the test requirements had been satisfied. Specifically,

Engineering Change 641099, which was included as part of Work Order (WO) 5515768 to demonstrate compliance with Technical Specification Surveillance Requirement (TS SR) 3.3.1.1.13 for Function 6, failed to incorporate the requirements and acceptance criteria contained in the applicable design documents associated with the TS SR. Additionally, the licensee failed to evaluate the test results documented in WO 5515768 to assure test requirements had been satisfied.

Failure to have Appropriate Procedure for Performing Reactor Protection System Main Steam Isolation Valve Closure Response Time Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-03 Open/Closed

[H.3] - Change Management 71111.21M The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have procedures appropriate to the circumstance for performing response time testing of the reactor protection system main steam isolation valve closure function, an activity affecting quality. As a result, Technical Specification Surveillance Requirement 3.3.1.1.17, which required measuring this response time, had been missed since October 1, 1993.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000461/2025010-04 10 CFR 50.59 Evaluation Needed to Determine if Standby Liquid Control Technical Specification Bases Change Required NRC Approval 71111.21M Open

INSPECTION SCOPES

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

REACTOR SAFETY

===71111.21M - Comprehensive Engineering Team Inspection The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:

Structures, Systems, and Components (SSCs) (IP section 03.01)===

For each component sample, the inspectors reviewed the licensing and design bases including:

(1) the Updated Final Safety Analysis Report (UFSAR);

(2) the Technical Specifications (TS); and

(3) the Operational Requirements Manual (ORM). The inspectors reviewed a sample of operating procedures (including normal, abnormal, and emergency procedures) and overall system/component health (including condition reports and operability evaluations, if any). The inspectors performed visual inspections of the accessible components to identify potential hazards and/or signs of degradation. Additional component-specific design attributes reviewed by the inspectors are listed below.

(1)125 Volt Direct Current (DC) Battery Charger 1C (1E22-S001E)

3. System Health (Failures, Condition Reports (CRs), operability evaluations

(OP Evals))

5. Electrical Design Calculations and Considerations:

a.

Sizing b.

Sizing of protective fuses/breakers/relays c.

Protective relays/breakers

(2) 4160 Volt Alternating Current (AC) Engineered Safety Feature (ESF) Bus 1C1 (1E22-S004)

6. Protection against External Events:

a.

Seismic b.

Fire

7. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Load testing b.

TS surveillances c.

Relay calibration d.

Terminal corrosion resistance

8. Electrical Design Calculations and Considerations:

a.

Loading calculations b.

Short circuit calculations c.

Voltage regulation d.

Coordination calculations e.

Bus capacity f.

Degraded voltage protection g.

Overcurrent protection h.

Loss of voltage i.

Harmonic content j.

Transmission system protection k.

Surge protection l.

Cable ampacity m. Protective devices and trip set points

(3) B Reserve Auxiliary Transformer (1AP02E)

5. Protection against External Events:

a.

Fire

6. Electrical Design Calculations and Considerations:

a.

Loading calculations b.

Short circuit calculations c.

Surge arrester protection d.

Auto transfer dynamic calculations e.

Degraded voltage protection f.

Loss of voltage g.

Breaker settings and ratings to prevent spurious tripping h.

Breaker control voltage i.

Grounding j.

Protective devices

7. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Load testing b.

TS surveillances c.

Terminal corrosion resistance

(4) Division 3 Diesel Generator (1DG01KC)

4. Protection against External Events:

a.

Flooding, including sump pump b.

Seismic

5. Mechanical Design Calculations and Considerations:

a.

Fuel oil transfer design (e.g., flow capacity, NPSH)b.

Combustion air supply design c.

Exhaust system design d.

Starting air design, including SBO recovery capability

6. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Fuel oil volume b.

Starting air

7. Electrical Design Calculations and Considerations:

a.

Loading b.

Control logic c.

Fuel oil transfer pump circuitry d.

Output breaker control logic e.

Protective relay setpoint f.

Capability to start under degraded voltage conditions

(5) Division 3 Shutdown Service Water Pump (1SX01PC)

6. Protection against External Events:

a.

Flooding, including sump pump b.

Seismic c.

HELB d.

Fire

7. Mechanical Design Calculations and Considerations:

a.

Flow capacity b.

Flow balance c.

Minimum flow d.

Required submergence (NPSH, vortexing)e.

Gas intrusion f.

Pump cooling g.

Room heat-up calculations h.

Room cooling

8. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Pump comprehensive in-service testing (IST) surveillances b.

Flow balance/capacity tests c.

Pump quarterly IST surveillances d.

Emergency water make-up

9. Electrical Design Calculations and Considerations:

a.

Pump motor brake horsepower b.

Control logic

(6) High Pressure Core Spray Pump (1E22-C001)

2. Protection against External Events:

a.

Flooding, including sump pump

3. Design Calculations and Considerations:

a.

Flow capacity b.

Minimum flow c.

Required submergence (net positive suction head, vortexing) d.

Water supply availability e.

Gas intrusion and accumulation

4. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Pump comprehensive IST surveillances b.

Pump quarterly IST surveillances

(7) Shutdown Service Water Pump Room 1C Coil Cabinet (1VH07SC)

6. Protection against External Events:

a.

Flooding, including sump pump b.

Seismic c.

Fire

7. Design Calculations and Considerations:

a.

Minimum cooling water flowrate b.

Maximum cooling water temperature c.

Minimum working fluid flowrate d.

Maximum working fluid temperature e.

Tube plugging limit f.

Fan motor minimum voltage g.

Heat transfer capacity

8. Test/Inspection Procedures, Acceptance Criteria, and Recent Results:

a.

Flowrates b.

Inspection or thermal performance test c.

Eddy current

Modifications (IP section 03.02) (4 Samples)

(1) Engineering Change (EC) 636666, Revision 0, Change 4KV Bus High and Low Voltage Alarm Setpoints for Buses Fed from RAT B or the ERAT

(2) EC 640929, Revision 0, Install Line Stop for 1SX062B Valve Replacement

(3) EC 641078, Revision 0, Substitute LS #3 with OPT ISOL for RPS Input from MSIV 1B21F022D

(4) EC 641099, Revision 0, Use of LS3 in Place of LS5 as Source of MSIV Closure Indication Input to RPS for 1B21F022D 10 CFR 50.59 Evaluations/Screening (IP section 03.03) (15 Samples)

(1) Evaluation CL-2021-E-020, GNF3 Impact on the Alternate Source Term Core Inventory, AST LOCA, FHA and CRDA

(2) Evaluation CL-2024-E-012, EC 641078 - Substitute LS #3 with Opt Isol for RPS Input from MSIV 1B21F022D

(3) Evaluation CL-2024-E-022, EC 641639 - Install a Backup Pressure Regulator for 1N66N516B Solenoid to Provide Manual Control Capability

(4) Screening CL-2022-S-011, Extend the Interval for ORM Logic System Functional Test of Feedwater / Main Turbine Trip on High Reactor Water (Level 8)

(5) Screening CL-2022-S-027, Emergency Diesel Generator Pyrometer Upgrade

(6) Screening CL-2022-S-101, EC 637534 - Pad Reinforcement for 1SX15AB-4" Min Wall Repair IAW N-789-3

(7) Screening CL-2023-S-015, 0FISWO013 Low Flow Trip Bypass for Chiller 0WO02CA and Chilled Water Pump 0WO03PA

(8) Screening CL-2023-S-025, EC 636711 - RWCU Differential Flow Modification

(9) Screening CL-2023-S-040, EC 639809 - Permanently Remove Motor Temperature Sensor 1TEVP080C in VP A Chiller

(10) Screening CL-2023-S-052, Leave a Temporary Sump Pump in Containment RF for Greater than 90 Days

(11) Screening CL-2024-S-004, EC 640439 - CPS RAT B LTC in Manual Tap Change Evaluation Support

(12) Screening CL-2024-S-032, Installation a Temporary Shield Over Sensors 1E31N559 A/B

(13) Screening CL-2024-S-010, TS SR 3.1.7.4 and SR 3.1.7.6 Bases Change

(14) Screening CL-2024-S-036, Revising 1WX020 and 1WX019 Closure Time

(15) Screening CL-2024-S-038, Surveillance Test Interval Change Request (STICR)

CL-19-006 Rev. 1 - Extend APRM Channel Functional Testing Frequency from Every 6 Months to Every 24 Months

Operating Experience Samples (IP section 03.04) (2 Samples)

(1)04712177, Emergency Diesel Generator Digital Reference Unit Issue

(2) 04744556, Review OPEX #56938 High Pressure Core Spray Failure

INSPECTION RESULTS

Failure to Perform Walkdowns Required by ASME Code Case N-789 Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-01 Open/Closed

[H.5] - Work Management 71111.21M The inspectors identified two examples of a finding of very low safety significance (Green)and an associated non-cited violation (NCV) of 10 CFR 50.55a(b)(5) for the failure to comply with the requirements of ASME Code Case N-789-3 and N-789-5. Specifically, the licensee failed to monitor pressure pads installed on ASME Code Class 3 piping for leakage every 31 days in accordance with ASME Code Case N-789.

Description:

In December 2021, the licensee performed an ultrasonic examination of ASME Code Class 3 piping in the shutdown service water (SX) system (piping section 1SX15AB-4") to evaluate potential thinning due to corrosion. The licensee discovered a section of piping had thinned due to microbiologically influenced corrosion (MIC) and evaluated the piping for continued service. At the time of discovery, the piping remained above the required ASME minimum wall thickness but would require repair before the next scheduled refueling outage to maintain structural integrity.

ASME Code Case N-789-3 allows the application of structural and pressure pads to areas of Class 2 and 3 moderate energy carbon steel raw water piping experiencing wall thinning from localized erosion, corrosion, cavitation, or pitting. On December 1, 2022, the licensee repaired the thinned section of piping using the alternate repair requirements of ASME Code Case N-789-3 by installing a pressure pad over the affected section of piping.

ASME Code Case N-789 also imposes an in-service monitoring requirement for installed pressure pads which consists of monitoring for evidence of leakage on a monthly (every 31 days) frequency. The licensee generated tasks under Work Order (WO) 5213314 to perform the required monthly leakage examinations.

The inspectors reviewed 50.59 Screening CL-2022-S-101 and EC 637534, which were written to support implementation of ASME Code Case N-789-3. On January 29, 2025, the inspectors identified the licensee had failed to complete one of the monthly monitoring activities within 31 days. Specifically, the licensee performed monthly monitoring on January 23, 2023. However, instead of performing the next monthly monitoring on, or before, February 23, 2023, the monthly monitoring activity was not completed until February 28, 2023 (36 days later). The inspectors determined all previous and remaining pressure tests were completed in accordance with code requirements.

Following identification of the original non-compliance with ASME code case requirements, the licensee performed an extent-of-condition review and identified an additional non-compliance with ASME Code Case N-789. On December 9, 2024, the licensee applied ASME Code Case N-789-5 to a separate section of thinned service water piping, 1SX14AA-3", to install a pressure pad and restore the section of piping to the original design thickness. Following installation, the licensee performed a post-installation pressure test on December 12, 2024, and returned the system to service. The licensee created an action in their work management system to schedule the remaining monthly leakage walkdowns, but this action was never assigned for scheduling. Consequently, the required monthly leakage walkdowns, which were required to begin by January 12, 2025, were not performed. After discovery on February 6, 2025, the licensee immediately performed a leakage walkdown, verified there was no leakage from the pressure pad installed on 1SX14AA-3", and scheduled work orders to perform monthly leakage walkdowns until the component is replaced.

Corrective Actions: The licensee performed a walkdown to examine 1SX15AB-4" for leakage on February 28, 2023, and verified no leakage was present. The licensee performed a walkdown to examine 1SX14AA-3" for leakage on February 6, 2025, and scheduled additional monthly examinations until the component is scheduled for replacement.

Corrective Action References: AR 4835310, CETI: Missed VT-2 for 1SX14AA Patch Weld Repair per N-789-5 and AR 4835447, CETI: VT-2 Not Completed Within 30 Day Required Frequency

Performance Assessment:

Performance Deficiency: The licensees failure to monitor ASME Code Class 3 components with installed pressure pads for leakage every 31 days was contrary to the requirements of ASME Code Case N-789 and 10 CFR 50.55(a)(b)(5), and was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to monitor ASME Code Class 3 components with installed pressure pads for leakage in accordance with ASME Code Case N-789 would have the potential to allow leakage from the affected component to occur unmonitored and challenge the structural integrity of the affected piping.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined this finding was of very low safety significance (Green), because they answered all Exhibit 2, Mitigating Systems Screening Questions, Mitigating SSCs and PRA Functionality questions no.

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, for both 1SX15AB-4" and 1SX14AA-3", the licensee failed to schedule walkdowns to monitor the affected components for leakage in accordance with the requirements of ASME Code Case N-789. Consequently, the required walkdowns were not performed within the required frequency.

Enforcement:

Violation: 10 CFR 50.55a(b)(5), Conditions on Inservice Code Cases, states, in part, that licensees may apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section, without NRC approval.

10 CFR 50.55(a)(3)(ii) states, in part, the use of code cases provided in NRC Regulatory Guide 1.147 is acceptable with the specified conditions in those guides when implementing the editions and addenda of the ASME BPV Code and ASME OM Code incorporated by reference in paragraph (a)(1) of this section.

10 CFR 50.55(a)(1)(ii)(C)(53) incorporates the 2013 Edition of the ASME Boiler Pressure Vessel Code by reference into 10 CFR 50.55a.

Clinton Power Station Inservice Inspection Program Plan - 4th Interval, Revision 0, establishes the Code of Record for the Fourth 10-Year Inservice Inspection Program as the ASME Boiler Pressure and Vessel Code, 2013 Edition, as incorporated by reference in 10 CFR 50.55a.

Regulatory Guide 1.147 incorporates ASME Code Case N-789-3 and ASME Code Case N-789-5 into Revisions 20 and 21, respectively.

ASME Code Cases N-789-3 and N-789-5 allow, in part, for areas of Class 2 and Class 3 moderate energy, carbon steel, raw water piping experiencing wall thinning from localized corrosion to be reinforced by the application of reinforcing pads to the surface of piping.

Additionally, both Code Cases require, in part, that installed pressure pads be monitored every 31 days for signs of leakage.

Contrary to the above, from February 23, 2023, to February 28, 2023, and from January 12, 2025, to February 6, 2025, the licensee failed apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of this section.

1) Specifically, the licensee failed to monitor ASME Code Class 3 component 1SX15AB-4, a moderate energy, carbon steel raw water pipe experiencing wall thinning from localized corrosion, for signs of leakage every 31 days after applying a reinforcing pad in accordance with ASME Code Case N-789-3.

2) Specifically, the licensee failed to monitor ASME Code Class 3 component 1SX14AA-3, a moderate energy, carbon steel raw water pipe experiencing wall thinning from localized corrosion, for signs of leakage every 31 days after applying a reinforcing pad in accordance with ASME Code Case N-789-5.

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

Failure to Perform Appropriate Testing of Main Steam Isolation Valve 1B21-F022D Reactor Protection System Instrumentation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-02 Open/Closed

[H.3] - Change Management 71111.21M The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure testing required to demonstrate the 1B21-F022D main steam isolation valve reactor protection system instrument would perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits contained in applicable design documents, and for the failure to evaluate the test results to assure the test requirements had been satisfied. Specifically, Engineering Change 641099, which was included as part of Work Order (WO) 5515768 to demonstrate compliance with Technical Specification Surveillance Requirement (TS SR) 3.3.1.1.13 for Function 6, failed to incorporate the requirements and acceptance criteria contained in the applicable design documents associated with the TS SR.

Additionally, the licensee failed to evaluate the test results documented in WO 5515768 to assure test requirements had been satisfied.

Description:

The reactor protection system (RPS) at Clinton Power Station was designed to initiate a reactor scram when one or more monitored parameters exceeded their specified limit to preserve the integrity of the fuel cladding and the reactor coolant system, and to minimize the energy that had to be absorbed following a loss of coolant accident. Main Steam Isolation Valve (MSIV) position was one of the monitored parameters. Specifically, a limit switch instrument mounted on each of the eight MSIVs signaled closure of its respective MSIV to the RPS. Each of the limit switches were arranged to open before its respective MSIV was more than 15 percent closed (the analytical limit) to provide the earliest positive indication of MSIV closure to the RPS. The MSIV RPS limit switch instruments were covered by TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation, which required the RPS instrumentation for each function in Table 3.3.1.1-1, Reactor Protection System Instrumentation, to be operable during the applicable modes specified in the table. Table 3.3.1.1-1, Function 6, Main Steam Isolation Valve-Closure, required the MSIV limit switches to be operable when the reactor was operating in Mode 1.

On March 10, 2024, while performing testing of the 1B21-F022D MSIV, the licensee discovered limit switch #5 (LS5), the RPS instrument on the valve, was not functioning. The licensee declared the Division 4 RPS instrument inoperable and entered limiting condition for operation (LCO) 3.3.1.1, Required Action A.1, which required placing the associated instrument channel of the affected function in trip within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or in accordance with the risk-informed completion time. Subsequently, the licensee completed modifications EC 641078 and EC 641099 to connect limit switch #3 (LS3), another limit switch on the 1B21-F022D MSIV, into the Division 4 RPS logic in lieu of the normally connected LS5.

LS3 was used as an installed spare since it was the same model number as LS5, was procured as safety related, and was installed on the MSIV at the same time as LS5. However, LS3 was not subjected to surveillance testing following initial installation because it only provided an input to a startup testing patch panel that was not normally monitored. After connecting LS3 into the Division 4 RPS logic, the licensee completed WO 5515768 on March 30, 2024, to restore the 1B21-F022D MSIV RPS instrument to an operable status.

Between January 27 and February 13, 2025, the inspectors reviewed TS Surveillance Requirement (SR) 3.3.1.1.13, Perform Channel Calibration, for Function 6 and noted the TS SR Allowable Value was an MSIV position of less than or equal to 13 percent closed. The inspectors determined note

(c) of Table 3.3.1.1-1 applied to TS SR 3.3.1.1.13 and it required the instrument channel setpoint to be reset to a value within the As-Left Tolerance of the actual trip setpoint as specified in the Operational Requirements Manual (ORM). Table 1, Reactor Protection System Instrumentation Trip Setpoints, of Attachment 2 of the ORM listed the Function 6 MSIV - Closure Actual Trip Setpoint as 7 percent closed. Table 17, Nominal Trip Setpoints, of Attachment 2 of the ORM listed the Function 6 Nominal Trip Setpoint (NTSP) as 7 percent and referred to design calculation IP-C-0075 as providing the As-Left Tolerance. Calculation IP-C-0075 listed the As-Left Tolerance as 5-9 percent closed.

The inspectors reviewed WO 5515768, Task 22 and noted it was credited to demonstrate compliance with TS SR 3.3.1.1.13 for the 1B21-F022D MSIV RPS instrument. However, the WO relied on EC 641099, Use of LS3 in Place of LS5 as a Source of MSIV Closure Indication Input to RPS for 1B21F022D, to demonstrate TS SR 3.3.1.1.13 was met instead of performing CPS 9431.10D, RPS Main Steamline Isolation Valve B21-F022D Channel Calibration. The inspectors noted surveillance procedure CPS 9431.10D could only be performed when the reactor was operating in Mode 4 (cold shutdown) or Mode 5 (refueling)because it required access to the MSIV in the drywell. The surveillance procedure required the licensee to physically measure the valve stroke with a ruler or tape measure capable of measuring 64ths of an inch to determine the precise valve position at which the RPS limit switch opened, thereby demonstrating TS SR 3.3.1.1.13 was met. The inspectors reviewed EC 641099 and noted that although the calculation included in the EC determined LS3 would provide an input to RPS at an MSIV position of 8.77 percent closed, there was a significant amount of uncertainty associated with the value. Specifically, the inspectors questioned the inputs and assumptions of the calculation which included using data from a static stroke of the 1BF022D MSIV performed in 2008 as an input into the calculation (MSIV slow closed stroke time) and an assumption that the static stroke time was unchanged although there was known variability in the stroke time and a significant amount of years had passed since the data was gathered. That data was then compared to a partial stroke of the valve during dynamic conditions under WO 5515768 to determine the valve position at which LS3 would provide the input into the RPS. In addition, the timing of the partial stroke of the valve was based on steam flow impacts because it was unknown when the MSIV actually started to move once it was demanded to slow close. The calculation also assumed the valve stroke speed was constant, although the motive force for the valve was from a spring which does not displace at a constant velocity. Overall, the licensee was unable to quantify the uncertainties created by the methodology used in EC 641099. Therefore, the inspectors questioned whether WO 5515768, Task 22 and EC 641099 incorporated the requirements and acceptance criteria contained in the applicable design documents associated with the TS SR and could be credited to demonstrate compliance with TS SR 3.3.1.1.13 after completing the modification to use LS3 as the RPS input for valve 1B21-F022D.

Corrective Actions: The licensee entered the inspectors concerns into the corrective action program as AR 4837051 on February 13, 2025. The licensee determined SR 3.3.1.1.13 was not met for the 1B21-F022D, Function 6 RPS instrument. The licensee invoked TS SR 3.0.3 which allowed them to delay entry into the required actions of LCO 3.3.1.1 for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever was greater, to allow performance of the surveillance. The licensee concluded the information in EC 641099 established a reasonable expectation the surveillance would be met when performed. The licensee also performed a risk evaluation per the requirements of SR 3.0.3 which concluded delaying the performance of the surveillance for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was low risk. The licensee planned to perform the surveillance at the next reasonable opportunity during the next refueling outage.

Corrective Action References: AR 4837051, CETI: MSIV F022D RPS MSIV-Closure Channel Calibration

Performance Assessment:

Performance Deficiency: The licensee failed to assure testing required to demonstrate the 1B21-F022D MSIV RPS instrument would perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits contained in applicable design documents. Additionally, the licensee failed to evaluate the test results to assure the test requirements had been satisfied, which was contrary to 10 CFR 50, Appendix B, Criterion XI, and a performance deficiency.

Specifically, EC 641099, which was included as part of WO 5515768 to demonstrate compliance with TS SR 3.3.1.1.13 for Function 6, failed to incorporate the requirements and acceptance criteria contained in the applicable design documents associated with the TS SR.

The licensee also failed to evaluate the test results documented in WO 5515768 to assure test requirements had been satisfied.

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 failure to appropriately test the RPS instrumentation did not ensure the availability, reliability, and capability of RPS to respond to a main steam isolation event to prevent undesirable consequences.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined this finding was of very low safety significance (Green) using Exhibit 2, Mitigating Systems Screening Questions, Section C, Reactor Protection System (RPS).

Specifically, although the finding affected a single RPS trip signal to initiate a reactor scram, the functions of other redundant trips or diverse methods of reactor shutdown were not affected.

Cross-Cutting Aspect: H.3 - Change Management: Leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, when deciding to credit EC 641099 to satisfy the requirements of TS SR 3.3.1.1.13 in lieu of performing the appropriate surveillance test, the licensee failed to use a systematic process for evaluating and implementing the change to ensure nuclear safety remained the overriding priority.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. It also requires that test results be documented and evaluated to assure that test requirements have been satisfied.

Contrary to the above, on March 30, 2024, the licensee failed to assure that testing required to demonstrate the 1B21-F022D MSIV RPS instrument would perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits contained in applicable design documents, and failed to evaluate the test results to assure the test requirements had been satisfied. Specifically, EC 641099, which was included as part of WO 5515768 to demonstrate compliance with TS SR 3.3.1.1.13 for Function 6, failed to incorporate the requirements and acceptance criteria contained in the applicable design documents associated with the TS SR.

Additionally, the licensee failed to evaluate the test results documented in WO 5515768 to assure test requirements had been satisfied.

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

Failure to have Appropriate Procedure for Performing Reactor Protection System Main Steam Isolation Valve Closure Response Time Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000461/2025010-03 Open/Closed

[H.3] - Change Management 71111.21M The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have procedures appropriate to the circumstance for performing response time testing of the reactor protection system main steam isolation valve closure function, an activity affecting quality. As a result, Technical Specification Surveillance Requirement 3.3.1.1.17, which required measuring this response time, had been missed since October 1, 1993.

Description:

Clinton Power Station Technical Specifications Section 1.1 defines RPS response time as the time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

On November 22, 2023, the response time definition was revised, via License Amendment 250, to include the following language:

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

While reviewing Procedure CPS 9431.20, RPS Main Steam Isolation Valve Response Time Test, on February 14, 2025, the inspectors noted the procedure failed to measure the time interval specified in the TS definition of response time. Instead, the licensee measured the RPS MSIV closure response time interval from the input of the RPS solid state logic to the de-energization of the scram pilot valve solenoids and then added an additional 10 milliseconds

(ms) to account for the time needed for the RPS MSIV closure parameter to exceed the RPS trip setpoint until the RPS signal reached the RPS solid state logic. Because the methodology provided in CPS 9431.20 did not meet the TS definition of response time for the MSIV closure function, the inspectors were concerned TS SR 3.3.1.1.17, which required response time testing of the RPS MSIV closure function, was not met.

During this inspection, the licensee documented their position regarding the inspectors concerns via a white paper. Within the white paper, the licensee acknowledged response time testing for specific RPS components was eliminated via the NRCs approval of NEDO-32291, System Analyses for Elimination of Selected Response Time Testing Requirements. However, the licensee also acknowledged the RPS MSIV closure limit switches were not among the specific components evaluated in NEDO-32291. The licensee concluded the methodology provided in CPS 9431.20 was appropriate and met the TS definition of response time and TS SR 3.3.1.1.17, because the methodology was based upon IEEE Standard 338-1977, IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems, which was endorsed (approved) by the NRC via Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, Revision 2.

The inspectors reviewed the licensees white paper and determined the licensees conclusion was incorrect. While the inspectors agreed the NRC had endorsed IEEE 338-1977 via Regulatory Guide 1.118, Revision 2, the revised TS definition of response time required both the component and the methodology to be reviewed and approved by the NRC. Additionally, the inspectors determined the revised RPS response time definition was based upon Technical Specification Task Force Traveler (TSTF) -332, ECCS Response Time Testing, Revisions 0 and 1. The inspectors reviewed TSTF-332, which Clinton Power Station implemented via License Amendment 250, and determined the TSTF was created to allow licensees to eliminate RPS response time testing for those components included in NEDO-32291 without having to submit a license amendment. License Amendment 250 also added words to the Clinton Power Station TS Bases, which states, When the requirements of Reference 10 (NEDO-32291) are not satisfied, sensor response time must be measured.

Since the MSIV closure limit switches were not included in NEDO-32291, no other license amendments had been submitted to the NRC to approve the components and methodology included in CPS 9431.20, and CPS 9431.20 failed to include steps to ensure the time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot solenoids was measured (including the sensor response time), the inspectors determined CPS 9431.20 was not appropriate to ensure RPS response time testing of the MSIV closure limit switches met the requirements of TS SR 3.3.1.1.17.

Corrective Actions: On April 29, 2025, the licensee invoked TS SR 3.0.3 due to not satisfying TS SR 3.3.1.1.17. This action allowed the licensee to delay entry into the required actions of TS 3.3.1.1 for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever was greater, to allow performance of the surveillance. The licensee concluded the successful performance of channel calibration and functional testing established a reasonable expectation the MSIV closure RPS response time testing surveillance would be met when performed. The licensee also performed a risk evaluation per the requirements of SR 3.0.3 which concluded delaying the performance of the surveillance for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was low risk. The licensee planned to submit a license amendment, requesting review and approval of the components and methodology outlined in the current version of CPS 9431.20 prior to the next refueling outage. If this review and approval is not received prior to the next refueling outage, the licensee planned to perform testing which met the TS RPS response time definition.

Corrective Action References: AR 4837054, CETI: MSIV Closure RPS Response Time Testing Methodology, and AR 4861577, CETI: Inadequate MSIV Closure RPS Response Time Testing

Performance Assessment:

Performance Deficiency: The licensees failure to have a procedure appropriate to the circumstance for testing the RPS MSIV closure limit switches to ensure TS SR 3.3.1.1.17 continued to be met was contrary to 10 CFR 50, Appendix B, Criterion V, and was a performance deficiency. Specifically, CPS 9431.20 failed to include steps to ensure the time interval from when the RPS MSIV closure limit switches exceeded their RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids was measured.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to have procedures for appropriately testing the RPS MSIV closure function response time reduces the assurance these components will appropriately respond to initiating events.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that this finding is of very low safety significance (Green) because they answered No to the Exhibit 2, section C, Reactor Protection System (RPS) screening question.

Cross-Cutting Aspect: H.3 - Change Management: Leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.

Although the performance deficiency occurred more than 3 years ago, the inspectors determined the cross-cutting aspect was reflective of present performance as provided in Section 3.14a of Inspection Manual Chapter 0612, Issue Screening, because the licensee had an opportunity to identify Procedure CPS 9431.20 did not appropriately test the RPS response time for the MSIV closure function when the NRC approved the revised TS response time definition in late 2023. When presented with this issue during the inspection, the licensee initially concluded the testing methodology in CPS 9431.20 was appropriate because other nuclear units within the Constellation fleet were using a similar/the same methodology. The licensee also documented their initial position in a white paper which failed to fully evaluate the entire licensing basis to ensure safety-related equipment was being tested as required.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.

Clinton Power Station Technical Specifications Section 1.1 defines RPS response time as the time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

Procedure CPS 9431.20, RPS Main Steam Isolation Valve Response Time Test, was the procedure used to perform RPS response time testing of the MSIV closure function, an activity affecting quality.

Contrary to the above, between October 1, 1993, and April 29, 2025, the licensee failed to have documented procedures appropriate to the circumstances for performing RPS response time testing of the MSIV closure function. Specifically, the procedure failed to include steps to ensure the time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids was measured as required by TS SR 3.3.1.1.17.

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

Unresolved Item (Open)10 CFR 50.59 Evaluation Needed to Determine if Standby Liquid Control Technical Specification Bases Change Required NRC Approval URI 05000461/2025010-04 71111.21M

Description:

On February 14, 2025, the inspectors identified an Unresolved Item (URI) associated with the licensees 50.59 Screening CL-2024-S-010, CPS TS Bases Revision for SR 3.1.7.6 (IR 4506683). On April 16, 2024, the licensee approved the screening in support of a change to the Standby Liquid Control (SLC) Systems TS Bases Surveillance Requirement.

Prior to the change the TS Bases read:

SR 3.1.7.6 verifies each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position from the control room. This is acceptable since the SLC System is a manually initiated system.

Following the change the TS Bases read (emphasis added to the change/added language):

SR 3.1.7.6 verifies each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system.

The associated TS SR 3.1.7.6 requires:

Verify each SLC subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position, or can be aligned to the correct position.

The licensees operating procedure 9015.01, Standby Liquid Control System Operability, required entering an 8-hour shutdown statement under ITS 3.1.7 due to closing manual isolation valves and filling the SLC test tank with water (refer to AR 04506683). By adding the or locally by a dedicated operator at the valve control to the TS Bases the licensee would potentially be able to credit operator actions in the field (i.e., inside containment) to reposition valves from a non-accident position into their accident position in order to credit the TS Surveillance met and the system to remain operable.

The inspectors noted UFSAR Section 9.3.5.4, Testing and Inspection Requirements, states, in part:

Testing of the SLC pumps without firing the explosive primers may be accomplished by the use of locally mounted control switches. Operation of a pump in the test mode does not prevent a manual initiation of boron injection from the control room, should it be required.

The licensee had established part of the intent of the change was to create better agreement between the sites TS Bases language and NUREG-1434, Volume 2, Standard Technical Specifications General Electric BWR/6 Plants. In addition, the licensee had stated the UFSAR had provided an implicit allowance to credit said operator actions in the field.

However, the inspectors did not agree with the reasoning used by the licensee. Specifically:

a.

The TS Bases (prior to the change) and the UFSAR had explicit language only referencing manual actions from the main control room; b.

Some of the SLC valves, which might now be credited with local operator manual actions, were non-safety related and/or not part of the Inservice Testing Program (IST). By adding said actions, these valves might now have new safety functions (i.e., reposition) during an accident; and c.

The SLC test tank is non-safety related and has not been evaluated to ensure it is seismically qualified when it is full of water.

Based on the concerns identified above the inspectors determined the TS Bases change constituted a change that adversely affected how the UFSAR described SSC design functions for the SLC system were performed and/or controlled. Therefore, the proposed TS Bases change required a 50.59 Evaluation to determine if the change could be made without NRC approval.

Planned Closure Actions: The inspectors need the licensee to complete the 50.59 Evaluation.

Once the licensee informs the inspectors the evaluation is completed, the inspectors can review the conclusions and determine the next steps required to complete the identification, assessment, and dispositioning of the issue.

Licensee Actions: At the time of the inspection, the licensee stated they had yet to implement any procedure changes or physical plant changes as a result of the 50.59 screening results. However, the licensee acknowledged explicitly allowing the SLC pump to remain operable with dedicated field operations is in contradiction to the exact wording of UFSAR Section 9.3.5.4. Therefore, the licensee will be evaluating the issue to determine if NRC approval was required before implementing the change. The licensee has entered this issue into the corrective action program.

Corrective Action References: IR 04835220, CETI: Insufficient 50.59 Review for TS 3.1.7 Bases Change; IR 01131985, 1C41A002: Clinton Review of LaSalle CDBI SLC Test Tank OPEX Licensee-Identified Non-Cited Violation 71111.21M 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: 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, requires, in part, that, throughout the service life of a boiling-water-cooled nuclear power facility, components that are classified as ASME BPV [boiler pressure vessel] Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of editions and addenda of the ASME BPV Code that become effective subsequent to editions specified in paragraphs (g)(2) and

(3) of this section to the extent practical within limitations of design, geometry, and materials of construction of the components.

Clinton Power Station Inservice Inspection (ISI) Program Plan - 4th Interval, Revision 0, establishes the Code of Record for the Fourth 10-Year ISI program interval as the ASME BPV Code, 2013 Edition, as incorporated by reference in 10 CFR 50.55a.

Section XI-2013 Edition, Paragraph IWA-4520(a), states, in part, welding or brazing areas and welded joints made for fabrication or installation of items by a Repair/Replacement Organization shall be examined in accordance with the Construction Code identified in the Repair/Replacement Plan.

The Construction Code of Record identified in the licensees Repair/Replacement Plan 5443230-17 is the ASME BPV Code, 1974 Edition Summer Addenda. ASME BPV Code, 1974 Edition Summer Addenda Section III ND-5240, Category D Vessel Weld Joints and Similar Joints in Piping, Pumps, and Valves, Paragraph ND-5242, Piping, Pumps, and Valves, states the requirements for weld joints similar to Category D weld joints shall be as given in ND-5212.

Paragraph ND-5212, Piping, Pumps, and Valves, states, longitudinal weld joints in piping, pumps, and valves greater than 4 inches nominal pipe size shall be examined by either the magnetic, liquid penetrant, or radiographic methods.

Contrary to the above, on June 7, 2024, the licensee failed to ensure that throughout the service life of a boiling-water-cooled nuclear power facility, components that are classified as ASME BPV Code Class 1, Class 2, and Class 3 met the requirements set forth in Section XI of editions and addenda of the ASME BPV Code that become effective subsequent to editions specified in paragraphs (g)(2) and

(3) of this section to the extent practical within limitations of design, geometry, and materials of construction of the components. Specifically, as part of Repair/Replacement Plan 5443230-17, the licensee failed to examine weld FW-1 on line 1SX017MB, a 12 inch safety-related pipe on the shutdown service water system by the performance of a magnetic, liquid penetrant, or radiographic examination as required by Paragraphs ND-5242 and ND5215 of the 2013 Edition of the ASME BPV Code Section XI for the current 10-Year ISI program interval Section XI-2013 Edition.

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

The inspectors determined the finding was of very low safety significance (Green) because they answered, No, to all of the questions included in Section C, Support System Initiators, of Exhibit 1, Initiating Events Screening Questions. Specifically, when the licensee performed the magnetic particle test, no indications were identified.

Corrective Action References: AR

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public disclosure.

• On May 1, 2025, the inspectors presented the Comprehensive Engineering Team Inspection results to D. Shelton, and other members of the licensee staff.

• On February 14, 2025, the inspectors presented the Comprehensive Engineering Team Inspection interim inspection results to A. Krukowski, and other members of the licensee staff.

• On March 20, 2025, the inspectors presented the Comprehensive Engineering Team Inspection additional interim inspection results to A. Krukowski, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

01HP08

Determination of NPSH for Pump 1E22C001 at Source

Switchover

1B

01HP09

ITS Surveillance Requirements for HPCS

01HP13

NPSH Calculation - HPCS Suction from Suppression Pool

(Licensing Basis)

01HP15

Development of HPCS Pump Curves (1E22C001) and

Comparison with System Resistance Curves for All Operating

Modes

01ME077

Calculations for Flooding - Safe Shutdown Analysis

01SX014

FSAR Table 9.2-3 Aux Heat Loads

25568

Foundation Loads for Shutdown Service Water Pump

25569

Shutdown Service Water Pump Mounting Structure

065-017

Summary Report Clinton GL89-13 Program Report

19-AI-81

CPS Appendix R Safe Shutdown Component Circuit Analysis

19-AJ-74

Class 1E Distribution Panel Loading Calculation

19-AK-06

Calculation for Auxiliary Power System Analysis

19-AK-13

Analysis of Load Flow Short Circuit and Motor Starting

19-AN-02

Calculation for 4160V ESS Bus Main Reserve Feed and

Medium Voltage Bus Relay Settings

19-AN-04

480V ESF Switchgear Breaker and Associated Upstream

Relay Settings

19-AN-04

480V ESF Breaker and Associated Upstream Relay Settings

13C

19-AN-09

4160V Division 3 ESF Bus 1C1 Motor Relay Settings

19-AN-17

Diesel Relay Settings Div. III

19-AN-19

Calculation for Functional Requirements for 1st and 2nd Level

Undervoltage Relays at 4kV Buses 1A1, 1B1, & 1C1

19-AN-20

Circuit Breaker Settings for 480v High Pressure Core Spray

(HPCS) Motor Control Center (MCC) 1E22S002

2F

19-AN-22

Circuit Breaker Settings for 480V Motor Control Centers

1D

19-AN-31

Relay Settings for RATs A, B & C

71111.21M

Calculations

19-AN-37

LTC Control Settings for RATs A, B, & C

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

19-AN-40

RAT B Open Phase Detection Limits

08/13/2020

19-AQ-02

Evaluation of Voltages for LOCA Block Motor Start

and 89-10 Program MOV Motors

19-D-49

Class 1E 125VDC System Analysis

19-G-01

Cable Ampacities in 40°C/ 50°C Cable Trays

3C10-0182-003

Containment Subcompartment Parameters for Environmental

Qualification of Equipment

3C10-0377-001

Containment Subcompartment Analyses - RWCU Line Break

3C10-0485-001

Internal Flooding Analysis

3C10-0487-001

Screen House Winter Temperature Steady State Operating

Conditions

3C10-118-001

NUMARC 87-00 Station Blackout Equipment List

3C10-1188-001

Blackout Equipment List

04/07/1989

ATD-0208

WS/SX Heat Exchanger Paths Residence Times

CQD-019893

Dynamic Qualification of American Air Filter Coolers

EPU-T0903

Extended Power Uprate Task T0903 Station Blackout

0D

IP-E-0032

Uncertainty Calculation for Div I, II, III 4.16kV Degraded

Voltage Relays

IP-M-0471

CPS Post Fire Safe Shutdown Criteria

6a

IP-M-0479

Hazards and Operability Analysis for the Cooling Water

Screen House and Associated Structures, Systems and

Components

IP-M-0486

Shutdown Service Water (SX) System Hydraulic Network

Analysis Model and Flow Balance Acceptance Criteria

IP-M-0605

Flow Velocities in U1 SX Pump Bay

0A

IP-M-0761

Evaluation of Vortex in the RCIC Storage Tank for HPCS and

RCIC Suction Lines

1A

IP-M-0806

NPSH and Vortexing Analysis for SX01PA/B/C

IP-S-0132

Acceptance Criteria for Allowable Sediment Depth (Siltation) in

the CW Screenhouse

2a

VH-31

Performance Evaluation of SSW Pump Rooms A, B, C

Cooling Coils Under SX Flow Acceptance Limits

VZ-45

SX Room Cooler Airflow Test Evaluation

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

04506683

SLC Pump Operability Tech Spec Change

06/21/2022

04506683

SLC Pump Operability Tech Spec Change

06/21/2022

04706428

Void Identified during UT Testing on HPCS Piping

10/02/2023

04708726

BWRT in Leakage at 22 gpm from RT Leakby

10/11/2023

04712876

Containment RE Sump B 1RE05PB Running Every

Seconds

10/26/2023

04769346

Recomb 1B Solenoid Drain Valve Buzzing

04/25/2024

04800686

1E22F001 PVT Frequency Adjustment

09/10/2024

4072491

Div 2 DG 16 Cyl Pyrometer Showing Inaccurate Readings

11/08/2017

22704

Erratic Readings Div 2 DG Cyl Exhaust Temps

04/03/2018

4176450

Results of 1SX01PC Packing Sleeve Inspection

09/24/2018

4359065

Large Leak Div I SX Strainer Piping Elbow

07/26/2020

25627

Action Tracking for 2021 NRC Open Phase Inspection

05/25/2021

4479946

HK: Division 3 SX Pump Lagging Needs Replaced

2/22/2022

4481604

NRC ID: Issues with Code Case Adherence

03/01/2022

4551444

Unexpected Trip of RAT SVC

01/31/2023

4675383

Leakage from Packing of D3 SX Pump Higher than Expected

05/04/2023

4706090

RAT B LTC not Performing as Expected

09/30/2023

4708616

Lessons Learned RAT B

10/04/2023

4708788

Trip of RAT 4538 Circuit Switcher

10/11/2023

4710692

MCR: 5008-5L Div 2 4KV Bus Low Voltage Unexpected Lit

10/18/2023

4712177

OE: Emergency Diesel Generator Digital Reference Unit Issue

10/24/2023

4714921

RAT MSC Trip

11/02/2023

4719639

Unexpected Annunciator 5007-5M 4 KV Bus High Voltage

11/26/2023

24285

RAT B LTC MSC Trip

2/18/2023

4736680

MCR Annunciator RAT Trouble 5010-6A

01/26/2024

4744556

IRIS Exp. ID #581395 - High Pressure Core Spray Pump

Minimum Flow Valve Would Not Stroke Closed

01/31/2024

4756744

South Bus Voltage Transient

03/09/2024

4756831

1B21-F022D MSIV Closure Failed to Trip RPS

03/10/2024

4769497

NRC ID: SX Walkdown Identified Items Housekeeping and

Lighting

04/25/2024

Corrective Action

Documents

4779691

Received MCR Annunciator 5007-5M 4 KV Bus High Voltage

06/10/2024

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

4780706

MT/PT NDE Needs Performed of FW-1 in WO 5219407-44

06/14/2024

4793938

Critical Spare Issued for 1SX01PC

08/14/2024

4795794

LR: Internal Coatings Inspection Required for 1SX01PC

08/21/2024

4796050

1WX020 PCIV Closing Stroke Time Exceeded Limiting Value

08/22/2024

4796276

Unexpected Annunciator 5007-5M 4 KV Bus High Voltage

08/22/2024

4796849

Lessons Learned 1SX01PC and Revisions Procedure

for MMD

08/26/2024

4797027

1SX01PC Packing Sleeve has Pitting

08/27/2024

4801997

4160V Bus 1B1 and 1C1 High Voltage During VC Shift

09/15/2024

4805133

Air Start Motor 1DG16MB Did Not Exhaust Oil Mist

09/28/2024

4817298

Extend Operation with All Vital Buses on One Power Supply

11/13/2024

0486380

CETI: EDG Pyrometer and Thermocouple UFSAR Change

Required in 50.59 Screening CL-2022-S-027

03/13/2025

4832678

CETI: Loose Insulation Found in Div 3 SX Pump Room

01/27/2025

4832943

CETI: Overhead Light Fixture Bulb Out in Div 3 Battery Room

01/28/2025

4832950

CETI: PCRA - 5130.04 Does Not Have Tie to 3215.01P001

01/28/2025

4832956

CETI: Level 3 OPEX Review AR 4341793-02 Improperly

Closed

01/28/2025

4832968

CETI: HPCS Pump Room Walkdown Housekeeping

01/28/2025

4833122

CETI: Ladders in Control Room Improperly Stored

01/29/2025

4833587

CETI: Div 3 Battery Charger Missing Bolts

01/30/2025

4833657

CETI: CW IA Suction Bay NAOCL Inject Heavily Corroded

01/30/2025

4833661

CETI: Degraded Lighting in Div III DG Fuel Storage Room

01/30/2025

4834494

CETI: 4200.01 SBO Action Not Identified as TCA

2/03/2025

4835220

CETI: Insufficient 50.59 Review for TS 3.1.7 Bases Change

2/06/2025

4835239

CETI: USAR 9.3.5.4 Contains a Flawed Description

2/06/2025

4835310

CETI: Missed VT-2 for 1SX14AA Patch Weld Repair

per N-789-5

2/06/2025

4835445

CETI: ORM Section 1.3 Needs to be Revised

2/07/2025

4835447

CETI: VT-2 Not Completed within 30 Day Required Frequency

2/07/2025

4836380

CETI: 2025 - USAR Change Not Identified in EC 636619

2/11/2025

4836396

CETI: NRC Observation during SBO Walkdowns

2/11/2025

Corrective Action

Documents

Resulting from

Inspection

4836518

CETI: NRC Observation on UFSAR PCIV Closure Time

2/11/2025

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Tables

4836631

CETI: Identified Procedure Use and Adherence Deviation

2/12/2025

4836795

CETI: Panel Doors in Control Room Improperly Stored

2/12/2025

4836801

CETI: UFSAR Contains Outdated Values

2/12/2025

4836946

CETI: UFSAR Change 2024-009 Under 50.59 Needs Revised

2/13/2025

4836996

CETI: PM Due Date Not Tracked Properly

2/13/2025

4837051

CETI: MSIV F022D RPS MSIV - Closure Channel Calibration

2/13/2025

4837054

CETI: MSIV Closure RPS Response Time Testing

Methodology

2/13/2025

4837277

CETI: Observation HPCS Pump Test Uncertainty

2/14/2025

4837278

CETI: NRC Observation on Vendor Manuals

2/14/2025

4837283

CETI: UFSAR Still Contains Multiple References of SVC

2/14/2025

4837357

CETI NRC Observation Maintaining Split Bus Configuration

2/14/2025

E02-1AP01, Sh. 1

SLD Part 1 CPS Unit 1

AF

E02-1AP03, Sh. 1

Electric Loading Diagram CPS Unit 1

AD

E02-1AP04, Sh. 1

4160V and 6900V SWGR Relay Settings CPS Unit 1

L

E02-1AP04, Sh.

4160 ESF Bus 1A1, 1B1, & 1C1 Relay Settings

Clinton Power Station Unit 1

M

E02-1AP04, Sh.

Diesel Generator 1A, 1B & 1C Relay Settings CPS Unit 1

AA

E02-1AP04, Sh.

Transformer Tap Settings

E

E02-1AP04, Sh. 3

480V & 4160V Relay Settings Clinton Power Station Unit 1

F

E02-1AP09, Sh. 1

Synchronizing Diagram Part 1

F

E02-1AP12

Relaying and Metering Diagram DG 1C

P

E02-1HP99, SH.

2

High Pressure Core Spray Div. 3 DG

N

E02-1HP99, Sh.

205

Schematic Diagram HPCS Diesel Generator

N

E02-1HP99, Sh.

209

HPCS 125VDC Distribution System

K

Drawings

E02-1HP99,

Sh.103

HPCS Power Supply System

N

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

E03-1E22-S004,

Sh. 7

4160V Bus 1C1 (1E22-S004) (Unit 101, 102, 103)

X

M05-1036, Sh. 2

P&ID Diesel Generator Fuel Oil System

T

M05-1040

Instrument Air Turbine BLDG (IA) Clinton Power Station Unit 1

Clinton, Illinois

Y

M05-1047

Containment Building Floor Drain System (RF) Clinton Power

Station Unit 1

M

M05-1047

Floor Drain System (RF) Clinton Power Station Unit 1, Sheet 1

L

M05-1074

Hight Pressure Core Spray Clinton Power Station Unit 1

AH

M05-1076

Reactor Water Clean -UP (RT) Clinton Power Station Unit 1

P

389026

RAT B Open Phase Detection

394754

RAT B Open Phase Detection Algorithm

2785

1SX01PC Design Life and Vendor Manual Update

27131

RAT B SVC Long Term Abandonment

28200

ERAT SVC Long Term Abandonment

630673

Div 3 SX Pump Replacement Design Study

2319

Div. 1 & 2 EDG Motor Operated Potentiometer (MOP)

Replacement with a Digital Reference Unit (DRU)

635782

Evaluate ORM TR 4.2.12.4 Logic System Functional Test

Frequency Change

636666

Change 4kV Bus High and Low Voltage Alarm Setpoints for

Buses Fed from RAT B or the ERAT

636711

RWCU Differential Flow Modification

637534

Pad Reinforcement for 1SX15AB-4" Min Wall

Repair IAW N-789-3

639809

Permanently Remove Motor Temperature Sensor 1TEVP080C

in VP A Chiller Motor

640439

CPS RAT B LTC in Manual Tap Change Evaluation Support

640929

Install Line Stop for 1SX062B Valve Replacement

641078

Substitute LS#3 with Optical Isolators for RPS Input from

MSIV 1B

Engineering

Changes

641099

Use of LS3 in Place of LS5 as a Source of MSIV Closure

Indication Input to RPS for 1B21F022D

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

2718

RAT MSC Trip 1A1 Bus Overvoltage Transient

EC 639971

Temporary Transfer Water from Containment RF to

Containment RE

EC 639971

Temporary Transfer Water from Containment RF to

Containment RE

EC 641639

Install a Backup Pressure Regulator for 1N66N516B Solenoid

to Provide Manual Control Capability

IP-C-0132

RWCU Differential Flow Setpoint Analysis for Detecting Large

Leaks

19-AI-14

Fast Transfer of ESF Buses 1A1 & 1B1 Between RAT and

ERAT

2G

19-AJ-72

Evaluation of Minimum Voltage Required at the Safety Related

Buses to Support the 120V Loads Fed from the MCC

Distribution Panels

09/18/2021

19-AN-41

ERAT and RATB Open Phase Detection LOCA Analysis

19-AN-42

RATB and ERAT MSC Banks Protective Relay Settings

01A

19-AS-06

RATB and ERAT MSC Banks Switching Study

19-AX-03

Circulating Current During Manual Source Transfer

(RAT to ERAT or Vice Versa)

01-A

3C10-0485-001

Internal Flooding Analysis

CL-2023-S-052

50.59 Screening - Temporary Transfer Water from

Containment RF to Containment RE

CL-2024-E-022

50.59 Evaluation - Install a Backup Pressure Regulator for

1N66N516B Solenoid to Provide Manual Control Capability

CL-2024-S-010

50.59 Screening - CPS TS Bases Revision for SR 3.1.7.6

(IR 4506683)

CL-2024-S-022

50.59 Screening - Install a Backup Pressure Regulator for

1N66N516B Solenoid to Provide Manual Control Capability

CL-MISC-029

Clinton Power Station Open Phase Condition Evaluation

IP-C-0053

4KV Auxiliary Buses Computer Alarm Setpoints

10/20/2022

IP-C-0131

4KV Auxiliary Bus Voltage Drift Analysis

10/20/2022

Engineering

Evaluations

NEDC-32974P

Clinton Power Station SAFER/GESTR-LOCA Analysis Basis

Documentation

01/10/2000

Miscellaneous

License Amendment Request

01/27/1995

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

AC & DC System Health Report

2/01/2025

TA Green, BWR Owners Group to NRC

2/10/1995

CPS RTT White Paper

2/20/2025

13A180

License Amendment Request

01/13/2023

24-15

Unit 1 Standing Order - 345 kV Voltage Control Strategy

with RAT B LTC in Manual

21A9236

Engine-generator for High Pressure Core Spray

2A3153, Sh.1

Specification, Reactor Protection System

2A3153AD, Sh.1

Specification, Reactor Protection System

CL-2022-S-027

Replace EDG Analog Pyrometers with Digital Pyrometers

CL-2023-S-015

0FISWO013 Low Flow Trip Bypass for Chiller 0WO02CA and

Chilled Water Pump 0WO03PA

CL-2024-E-12

Substitute LS#3 with Opt. Isol. for RPS Input

from MSIV 1B21F022D

CL-2024-S-032

Temp Shield Over Sensors

CL-2024-S-036

1XW019 and 1WX020 Closure Time Revision

CL-2024-S-038

Surveillance Test Interval Change

CL-SURV-015

Risk Evaluation of Deficient Technical Specification

Surveillance for SR 3.3.1.1.13 for Div 4 RPS MSIV Closure

DC-SX-01-CP

Shutdown Service Water System Design Criteria

K2828-B

Shutdown Service Water Pump 1SX01PC

K2902

Special Coil Cabinets

PMC-22-137172

Create Periodic Replacement PM for New DRU Installed in

Div. 1 DG

08/29/2023

PMC-23-145568

Create Periodic Replacement PM for New DRU Installed in

Div. 2 DG

04/25/2024

RSS420

Operating Instructions, Three Phase Six Pulse ARR Series

SCR Float Charger

VPF 3842-002

HPCS Motor Curve

Operability

Evaluations

23827-02

Open Phase Operability Evaluation

Procedures

1014.11

6900/ 4160/ 480V Switchgear/ Circuit Breaker Operability

Program

5e

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

1019.05

Transient Equipment/Materials

26c

3001.01C001

Preparation for Startup Checklist

19f

211.01

Shutdown Service Water (SX)

33D

215.01P001

Placing Backup Pressure Air Regulator in Service

3309.01

High Pressure Core Spray (HPCS)

18d

3501.01

High Voltage Auxiliary Power System

3501.01D001

Monitoring Safety Related 4.16kV Bus Voltage Data Sheet

4c

3503.01

Battery and DC Distribution

3505.01

345 & 138kV Switchyard (SY)

3505.01C005

RAT B - LTC Manual Switching Order

1a

3514.01

25VDC Div. 3 Outage

3514.01C007

4160V Bus 1C1 (1E22-S004) Outage

3b

200.01

Loss of AC Power

27a

200.01

Loss of AC Power

200.01C001

MCR Cooling During a SBO

4c

200.01D010

Emergency Bus 1C1 Trip Data Sheet

5007.05

Alarm Panel 5007 Annunciators - Row 5

31a

5009.03

Alarm Panel 5009 Annunciators - Row 3

30d

5064.01

Alarm Panel 5064 Annunciators - Row 1

31C

5064.02

Alarm Panel 5064 Annunciators - Row 2

2A

8410.06

General Electric 4160 Power Circuit Breaker Maintenance

8410.08

General Electric 4160V Power Circuit Breaker Switchgear

(1E22-S004) Maintenance

4a

8410.30

Division III 4.16KV GE Magne-Blast Breaker Exchange

8451.05

Corrective Maintenance for Limitorque SMB-000, SMB-00,

and SB-00 Operators

11a

8451.06

Corrective Maintenance for Limitorque SMB-0 Through SMB-4

and SB-0, SB-1 and SB-3 Operators

9b

8451.08

Corrective Maintenance for Limitorque SMB-5T Operator

4a

9031.10

RPS Main Steam Isolation Valve Channel Functional

9080.03

Diesel Generator 1C Operability - Manual and Quick Start

Operability

9080.27

Unit Power Supply Manual Transfer Operability

1d

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

9082.01

Offsite Source Power Verification

2c

9431.10

RPS MSIV Calibration (Loop D)

9431.10D

RPS Main Steamline Isolation Valve B21-F022D Channel

Calibration

0a

9431.20

RPS Main Steam Isolation Valve Response Time Test

9431.20

RPS Main Steam Isolation Valve Response Time Test

CPS

215.01P001

Placing Backup Pressure Air Regulator in Service

CPS 3309.01

High Pressure Core Spray (HPCS)

18d

CPS 5130.04

Alarm Panel 5130 Annunciators - Row 4

27c

CPS 5130.04

Alarm Panel 5130 Annunciators - Row 4

CPS 9015.01

Standby Liquid Control System Operability

CPS 9051.01

HPCS Pump & HPCS Water Leg Pump Operability

OP-AA-108-117

Protected Equipment Program

OP-CL-102-106

Operator Response Time Master List at CPS

04884950-01

MM 1SD1-29 Lube/Inspect Door, Inspect Seal/Replace as

Needed

10/06/2020

04884951-01

1SD1-28 Lube/Visual Inspect Door Inspect Seal

05/08/2020

05088861

9051.01, 1E22C003 Comprehensive Pump Test

03/08/2022

293064

9051.01 1E22C003 Comprehensive Pump Test

06/15/2024

05582066

9051.01R22 HPCS Pump & WTR Leg Pump OPER

(RCIC STRG Tank)

2/05/2024

0772548

1B21F022D: EQ - Replace MSIV Actuator with New/Rebuilt

01/27/2008

1105807

EQ-CL013B Replace Limit Switches - 1B21F022D

10/22/2013

1686093-01

Replace Qualitrol 63X-1 and 63X-2 Relays 1AP02EB

05/06/2015

1686137-01

RAT B LTC Overhaul Inspection

05/10/2015

1765922

Divers Inspect/Clean Screenhouse Structure/GL 89-13

Program

2/12/2016

1787894-01

Replace 480 VAC Control Cabinet MCC Breakers - 1AP02EB

05/12/2017

1880203-01

Test Div III Bus Res Feed BKR Protective Relays

03/14/2018

1940985-01

Test/ Calibrate DG 1C Sync-Check Relay

10/04/2022

4596872

Replace 1SX01PC with New Design EC 404025

04/04/2016

Work Orders

4711157-01

Doble Test RAT B - 1AP02EB

10/12/2020

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

4782517

EOID: Valve Stem Broken for 1E22-F012

05/12/2018

4794542

Perform CPS9431.20 for RPS MSIV RTT

09/22/2019

4796857-01

Clean Insp Burnish HMA Relays 27SX1, 27SX2, 27SY1,

27SY2, 74-F1

09/25/2019

4796858-01

1E22S004 Perform Switchgear Maintenance IAW 8410.08

10/14/2019

4807687

Lube Inspect Door 1SD1-13

03/08/2022

4808234

Inspect 1SX01PC Throttle Shaft Sleeve for Corrosion

09/24/2018

4815807

C1R19 - NDES/GE Perform Inspection on Component

Supports

09/09/2022

4817490

9080.23R20 OP DG 1C Integrated Test

03/11/2023

4833619

1VH03A Set Up Chemical Inspect, Boroscope, Clean,

Eddy Current

06/09/2020

4834503

1SX01PC Rebuild Spare Pump

11/21/2018

4834624

Pull Entire Div 3 SX Pump and Inspect Pump Internals

06/06/2019

4839627

1SX01PC Inspect Packing Sleeve for Corrosion

11/29/2018

4844076

1SX01PC Inspect Packing Sleeve for Corrosion

06/06/2019

4859493-01

Replace Load Tap Changer (LTC) Controller for RAT B

09/23/2019

4886551-01

1AP02EB - Insulating Oil Tests

09/22/2020

24413

1SD1-12, Lube Inspect Door

03/16/2022

27032

1SD1-10 Lube Inspect Door

09/18/2020

4967576

MSL D Inboard Isolation Valve Indication Issue

10/09/2019

4981115-01

Test Fault Pressure Relays (2) Main Tank RAT B (NERC)

09/18/2023

4985706

9080.21R20 OP DG 1A Integrated Test (Except Section 8.5)

10/15/2021

4992375

1SX01PC Pull Entire Div 3 SX Pump and Inspect Internals

09/03/2021

4992526

RPS MSIV CC (Loop D)

10/12/2021

4992526-01

IM F022D 9431.10D (D20) CC *RPS MSIV CC (Loop D)

10/17/2021

28143

SWAP Breaker for 1E22C001 HPCS Pump, Relay Functional

Test

09/03/2024

29338

Perform UT on 1SX15AB-4"

01/07/2022

5108616-02

Clean and Inspect Div III Generator and Exciter

10/31/2022

213314

1SX15AB UT Data Lower than Expected

01/23/2023

214713

9069.01C20 OP SX Pump Oper Test

03/09/2022

215087

Pull Entire Div 3 SX Pump and Inspect Pump Internals

07/16/2024

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

219407

Stopple-Install Stopples for FC HX Isolation-SX

06/18/2024

260636-01

EOID 1AP02EB-TM8 Serveron Helium Gas Pressure Low

05/23/2022

292112

Perform CPS9031.10 for RPS MSIV

03/10/2024

297098

Replace Voltage Reg R3 Rheostat

10/07/2024

5306980-01

EMD 1DG01KC Repair Generator Output Terminations

11/22/2022

5307362-02

Unexpected 5062-2A Auto Trip 4160V Bus 1C1 Res Bkr

05/18/2023

21133

Obtain Airflow Measurements for Room

10/30/2024

21135

Vacuum Cooling Fins and/or Clean with Water

10/03/2024

21393

Test HX Performance per CPS 2602.01

11/01/2024

5344132-01

Preform an Operating Check of the Transformer Cooling

System

03/18/2024

5349016-02

Visual Inspect RAT B Radiators/ Fans

03/18/2024

5364314-02

Clean and Inspect Div III Generator and Exciter

09/30/2024

5364822

Div. 3 Diesel Generator Voltage Adjustment

10/05/2024

5403253-02

Anomalies Encountered During RAT Functional Testing

09/30/2023

5409932-01

Furan Oil Test - 1AP02EB

10/15/2024

5515768

MSL D Inboard Isolation Valve Issue

03/30/2024

5515768

MSL D Inboard Isolation Valve Indication Issue

03/30/2024

5539476

Pull Entire Div 3 SX Pump & Inspect Pump Internals

05/31/2024

5545933-01

9061.03A08 OP CT/DW Iso. Vlv. Op. (WX, WK 8-Timing)

08/29/2024

5612718

EDG 1C Monthly Test

01/28/2025

20004

9082.01R20 VER Offsite Source Power Verification

2/02/2025

22029

9080.01R20 VER Offsite Source Power Verification

2/09/2025