IR 05000336/2024002

Integrated Inspection Report 05000336/2024002 and 05000423/2024002

ML24226A378
Person / Time
Site:	Millstone
Issue date:	08/13/2024
From:	Matt Young NRC/RGN-I/DORS
To:	Carr E Dominion Energy
References
EA-24-073 IR 2024002
Download: ML24226A378 (1)

Text

August 13, 2024

SUBJECT:

MILLSTONE POWER STATION, UNITS 2 AND 3 - INTEGRATED INSPECTION REPORT 05000336/2024002 AND 05000423/2024002

Dear Eric Carr:

On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Millstone Power Station, Units 2 and 3. On July 18, 2024, the NRC inspectors discussed the results of this inspection with Michael O'Connor, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Four 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.

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

Document Control Desk, Washington, DC 20555- 0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Millstone Power Station, Units 2 and 3.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555- 0001; with copies to the Regional Administrator, Region I; and the NRC Resident Inspector at Millstone Power Station, Units 2 and 3. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading- rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Matt R. Young, Chief Projects Branch 2 Division of Operating Reactor Safety

Docket Nos. 05000336 and 05000423 License Nos. DPR- 65 and NPF- 49

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000336 and 05000423

License Numbers: DPR- 65 and NPF- 49

Report Numbers: 05000336/2024002 and 05000423/2024002

Enterprise Identifier: I-2024-002- 0038

Licensee: Dominion Energy Nuclear Connecticut, Inc.

Facility: Millstone Power Station, Units 2 and 3

Location: Waterford, CT

Inspection Dates: April 1, 2024 to June 30, 2024

Inspectors: J. Fuller, Senior Resident Inspector D. Antonangeli, Resident Inspector E. Bousquet, Resident Inspector D. McHugh, Reactor Inspector D. Werkheiser, Senior Reactor Analyst

Approved By: Matt R. Young, Chief Projects Branch 2 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 Millstone Power Station, Units 2 and 3, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Unit 3 Pressurizer Power-Operated Relief Valve Failed to Open Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.7] - 71111.15 Systems NCV 05000423/2024001-02 Documentation Closed EA-24-073 A finding of very low safety significance (Green) and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion VIII,

Identification and Control of Materials, Parts, and Components, was self-revealed on October 20, 2023, when the Unit 3 'B' pressurizer power-operated relief valve (PORV) failed to stroke open during surveillance testing. The licensees program for the identification and control of materials, parts, and components failed to prevent the installation and use of an incorrect solenoid- operated valve (SOV) on the 'B' pressurizer PORV. Specifically, the licensee installed an SOV that contained Stellite internal parts, which were known to be susceptible to leakage when exposed to steam and contributed to the failure of the 'B' PORV to open during surveillance testing.

Failure to Establish and Implement Appropriate Procedure/Instructions for Maintenance on the Unit 2 Control Room Heating and Ventilation System Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.5] - 71111.24 NCV 05000336/2024002-01 Operating Open/Closed Experience A finding of low safety significance (Green) and associated NCV of 10 CFR Part 50,

Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to incorporate appropriate acceptance criteria for refilling the refrigerant charge of the Unit 2 'A' control room air conditioning (CRAC) compressor within work instruction and procedures. Consequently, an incorrect amount of refrigerant was added which resulted in the inoperability of the 'A' train of the control room emergency ventilation system for longer than its technical specification allowed outage time.

Failure to Perform Vibration Measurements in Accordance with American Society of Mechanical Engineers Operation and Maintenance of Nuclear Power Plants Code Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.7] - 71111.24 Systems NCV 05000336, 05000423/2024002- 02 Documentation Open/Closed The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50.55a(f)(4)(ii) when the licensee failed to perform inservice testing on safety-related pumps in accordance with the American Society of Mechanical Engineers Operation (ASME) Operation and Maintenance (OM) Code. Specifically, the licensee failed to meet subsections ISTB3510(c), ISTB -3540(a), ISTB3540(d), and ISTB -3300(f) when it failed to take vibration measurements in appropriate locations on the pump bearing housings and did not take measurements in the same location for each test. This resulted in nonconservative data recorded during the surveillance tests.

Failure to Correct a Degraded Condition on the Main Generator Output Breaker Compressed Air System Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.11] - 71152A FIN 05000423/2024002- 03 Challenge the Open/Closed Unknown A self-revealed finding (FIN) of very low safety significance (Green) was identified for failure to correct a degraded condition on the main generator output breaker (MGOB) compressed air system in accordance with procedure MP-PROC- 000- PI-AA-200, Corrective Action. As a result, a moisture and oil buildup occurred in the 'B' phase supply pole air channel. This caused a ground fault on the 'B' phase of the MGOB. The protection relaying actuation resulted in a turbine trip and a reactor trip on May 30, 2023.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000336/2024-001-00 LER 2024-001-00 for 71153 Closed Millstone Power Station,

Unit 2, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting i n a Condition Prohibited by Technical Specifications LER 05000423/2023-001-00 LER 2023-001-00 for 71153 Closed Millstone Power Station,

Unit 3, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault

PLANT STATUS

Units 2 and 3 operated at or near rated thermal power for the entire inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp- manual/inspection-procedure/index.html.

Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase.

The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, conducted routine reviews using IP 71152, Problem Identification and Resolution, 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

Impending Severe Weather (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather on April 3, 2024.

71111.04 - Equipment Alignment

Partial Walkdown (IP Section 03.01) (3 Samples)

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

(1) Unit 2 charging system, including the 'B' charging pump from the suction piping below the volume control tank to the charging pump, on May 6 and 7, 2024

(2) Unit 3 'A' train safety injection pump cooling water piping on May 9, 2024

(3) Unit 3 accessible portions of the 'B' auxiliary feedwater (AFW) system piping from the demineralized water storage tank to the containment wall on May 28, 2024

Complete Walkdown (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit 3 main feedwater system on May 24, 2024, and June 11, 2024.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection (IP Section 03.01) (7 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) Unit 2 general area and reactor building closed-cooling water heat exchanger area (fire areas A-1A and A-1B) on April 24, 2024

(2) Unit 3 west service water cubicle (fire area CSW-4) on April 30, 2024

(3) Unit 2 containment recirculation valve room (fire area A-8B) on May 6 and 7, 2024

(4) Unit 3 north and south residual heat removal heat exchanger cubicles on the 21'-6" elevation (fire areas ESF-3 and ESF-6) during hot work activities on May 10, 2024

(5) Unit 2 cable vault on the 25'-6" elevation (fire area A-24) on May 21, 2024

(6) Unit 2 motor-driven AFW pump pit (fire area T-3) on May 21, 2024

(7) Unit 3 station blackout emergency diesel enclosure (fire area SBO-1) on May 22, 2024

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (2 Samples)

The inspectors evaluated readiness and performance of:

(1) Unit 3 'C' reactor plant component cooling water heat exchanger on June 21, 2024

(2) Unit 2 'A' emergency diesel generator air cooling heat exchanger, X83A, cooled by service water on June 27, 2024 (work order (WO)53203385496, WO53203313300)

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)

(1) The inspectors observed and evaluated licensed operator performance in the Unit 3 control room during their response to an earthquake on April 5, 2024, and during condenser backwashing activities on June 24, 2024.

(2) The inspectors observed and evaluated licensed operator performance in the Unit 2 control room during their response to an earthquake on April 5, 2024, and during restoration of a switchyard breaker on June 27, 2024.

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

(1) The inspectors observed and evaluated operator requalification training in the Unit 3 simulator on April 30, 2024.

(2) The inspectors observed and evaluated operator requalification training in the Unit 2 simulator on May 7, 2024.

71111.12 - Maintenance Effectiveness

Aging Management (IP Section 03.03) (1 Sample)

The inspectors evaluated the effectiveness of the aging management program for the following structures, systems, and components that did not meet their inspection or test acceptance criteria:

(1) Implementation of the Unit 2 buried piping inspection license renewal aging management program for the AFW piping between the condensate storage tank and turbine building contained within the condensate storage tank pipe trench on May 29, 2024

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management (IP Section 03.01) (6 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Unit 2 high risk plan and associated mitigating actions for emergent work on breaker 15G-8T-2 on April 10, 2024 (condition report (CR)1255840)

(2) Unit 3 elevated risk and associated mitigating actions described in the medium risk plan and the plant technical specifications for application of the 14-day allowed outage time during a planned extended emergency diesel generator maintenance outage from April 5 to 12, 2024

(3) Unit 2 elevated risk and mitigating actions associated with the planned maintenance outage for the 2A emergency diesel generator on June 3, 2024

(4) Unit 2 elevated risk and mitigating actions associated with the planned maintenance outage for the 2B emergency diesel generator on June 7, 2024

(5) Unit 3 increased risk and associated mitigating action associated with removing battery bus 6 from service to repair a battery cell leak on June 22, 2024 (CR1262569, WO53203422974)

(6) Unit 3 elevated risk and mitigating actions associated with the surveillance testing of the turbine-driven AFW pump and associated relay testing on June 26, 2024

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)

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

(1) Unit 3 'A' emergency diesel generator thrust bearing seal leak on April 10, 2024 (CR1256120)

(2) Unit 2 CRAC compressor not functioning properly on April 7 and 14, 2024 (CR1255747, CR1256509)

(3) Unit 2 'A' reactor building closed-cooling water pump with elevated iron identified in the inboard bearing oil along with a known out-of-tolerance clearance on the inboard bearing outer race on April 26, 2024 (CR1257537, CR1237468)

(4) Unit 3 structural integrity and operability evaluations for the 'A' safety injection pump cooling heat exchanger degraded pipe support on May 7, 2024 (CR1258795)

(5) Unit 2 reactor coolant system (RCS) loop 2B cold leg temperature instrument (T122CB) on May 19, 2024 (CR1259808, CR1247690)

(6) Unit 2 operability determination and supporting structural integrity evaluation for the

'C' service water strainer shell and nozzle base material less than code allowable on May 28, 2024 (CR1260581, CR1260512)

(7) Unit 3 'B' train reactor plant component cooling water containment isolation valve 3CCP*MOV45B after repairs on June 26, 2024 (CR1262615, WO53203423015)

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)

The inspectors evaluated the following temporary or permanent modifications:

(1) Unit 2 'C' high-pressure safety injection pump coupling design permanent modification change on May 14, 2024 (WO53102752180)

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) (8 Samples)

(1) Unit 3 'A' emergency diesel generator following extended maintenance outage on April 10, 2024 (WO53203363526)

(2) Unit 2 'A' CRAC system on April 9 and 16, 2024 (WO53203413663, WO53203414003)

(3) Unit 3 'A' safety injection pump cooling heat exchanger after repair was made to a degraded pipe support on May 10, 2024 (CR1258795, WO53203416414, WO53203417223)

(4) Unit 3 control building chilled water float valve inspection and replacement and strainer cleaning on May 10, 2024 (WO53203313725)

(5) Unit 2 'C' service water strainer after planned overhaul on May 31, 2024 (WO53203342298)

(6) Unit 3 'C' reactor plant component cooling water heat exchanger after planned inspections and cleaning on June 23, 2024 (WO53203406923)

(7) Unit 2 3-year instrument loop calibration for letdown heat exchanger flow transmitter (M2F-202) on June 24, 2024 (WO53203314667)

(8) Unit 3 'B' train reactor plant component cooling water containment isolation valve 3CCP*MOV45B after repairs on June 26, 2024 (CR1262615, WO53203423015)

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) Unit 2 'A' reactor building closed-cooling water system pump surveillance test on April 8, 2024

(2) Unit 3 'B' emergency diesel generator operability test on May 21, 2024

(3) Unit 3 turbine-driven AFW pump inservice testing on June 26, 2024

Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) Unit 3 'B' motor-driven AFW pump on April 22, 2024

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

(1) Units 2 and 3 triennial functional testing of beyond design basis equipment for one 480-volt emergency diesel generator and one 120- volt emergency generator on June 10, 2024

71114.06 - Drill Evaluation

Required Emergency Preparedness Drill (1 Sample)

(1) Unit 2 hostile action-based emergency planning rehearsal drill on May 1,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours (IP Section 02.01)===

(1) Unit 2, April 1, 2023 through March 31, 2024

(2) Unit 3, April 1, 2023 through March 31, 2024

IE03: Unplanned Power Changes per 7000 Critical Hours (IP Section 02.02) (2 Samples)

(1) Unit 2, April 1, 2023 through March 31, 2024

(2) Unit 3, April 1, 2023 through March 31, 2024

IE04: Unplanned Scrams with Complications (USwC) (IP Section 02.03) (2 Samples)

(1) Unit 2, April 1, 2023 through March 31, 2024

(2) Unit 3, April 1, 2023 through March 31, 2024

===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) Unit 3 automatic reactor trip caused by a main generator ground fault on May 30, 2023 (CA11902703)

(2) Unit 3 NRC-identified NCV 05000423/2023004-04, Failure to Identify and Correct a Degraded Overspeed Trip Mechanism on the 3A Emergency Diesel Generator (CA12282161)

71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)

(1) The inspectors reviewed the licensees corrective action program to identify potential adverse trends for active Units 2 and 3 service water strainer packing leakage that might be indicative of a more significant safety issue.

===71153 - Follow-up of Events and Notices of Enforcement Discretion

Event Report (IP Section 03.02)===

The inspectors evaluated the following licensee s event reporting determinations to ensure it complied with reporting requirements.

(1) Licensee Event Report (LER) 05000423/2023-001-00, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23208A287): The inspection conclusions associated with this LER are documented in this report under the Inspection Results Section 71152A. This LER is closed.

(2) LER 05000336/2024-001-00, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications (ADAMS Accession No. ML24162A088): The inspection conclusions associated with this LER are documented in this report under the Inspection Results Section 71111.24. This LER is closed.

INSPECTION RESULTS

Unit 3 Pressurizer Power-Operated Relief Valve Failed to Open Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.7] - 71111.15 Systems NCV 05000423/2024001 - 02Documentation Closed EA- 24- 073 A finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components, was self - revealed on October 20, 2023, when the Unit 3 'B' PORV failed to stroke open during surveillance testing. The licensee s program for the identification and control of materials, parts, and components failed to prevent the installation and use of an incorrect SOV on the 'B' pressurizer PORV. Specifically, the licensee installed an SOV that contained Stellite internal parts, which were known to be susceptible to leakage when exposed to steam and contributed to the failure of the 'B' PORV to open dur ing surveillance testing.

Description:

On October 20, 2023, with Millstone Unit 3 at 0 percent reactor power and in Mode 4, the 'B' pressurizer PORV, 3RCS*PCV456, failed to open during performance of SP 3601B.2, R eactor Coolant System Vent Path Operability Check, which was required prior to crediting the PORV for the cold over pressure protection system.

The pressurizer PORVs are solenoid actuated, pilot operated valves. The valves are part of the ASME Class 1 RCS pressure boundary and are required to open to provide pressure control during a safety grade cold shutdown. The valves are also credited for the cold overpressure protection system, a feed line break, loss of main turbine trip, loss of normal feedwater, and other inadvertent transients.

During the last operating cycle, the 'B' PORV began to leak on August 17, 2022. The leakage increased over the next 2 days, and on August 19, 2022, the licensee closed the associated block valve. However, the block valve did not fully isolate the PORV, and steam leakage continued throughout the operating cycle. At times during the cycle, leakage past the block valve and through the PORV SOV was as high as 245 gallons per day (0.17 gallons per minute), which was below the 10 gallons per minute technical specification limit for identified RCS leakage.

The licensee determined that the direct cause of the PORV failure to open was steam erosion of the override assembly top stem and valve ball that prevented the pilot SOV from lifting to provide the main body PORV the required differential pressure to stroke.

The licensees cause evaluation determined that the damage to the SOV internals was caused by the combination of the use of Stellite valve internals and the inability of the upstream pressurizer block valve to fully isolate the PORV. The Stellite valve internals were subjected to this steam cutting environment from August 17, 2022, to October 20, 2023 (429 days).

Station operating experience in a Millstone causal investigation performed in 2000 (A/R 00011126) documented that the Stellite internals utilized in the original pilot SOV design were inadequate. The original equipment manufacturer, Crosby, suspected that the prevalence of pilot SOV leakage within the industry was caused by the Stellite valve ball and seat materials being originally selected to be operated with water as a medium. They stated that the Stellite internals were susceptible to leakage when exposed to steam. In 2001, Westinghouse sent letter NEU- 01- 501 to the licensee offering a program to refurbish pilot SOVs by replacing the original ball and upper seat of the pilot assembly, which are made of Stellite 6B, with Inconel 718.

The licensee accepted Westinghouses offer to refurbish the SOVs. The licensee approved design change notice DM3- 00-0472- 01, Refurbished Solenoid Valve Assemblies Applicable to the MP3 Solenoid Power-Operated Relief Valves, on June 12, 2002, which changed the material of the ball and upper seat materials from Stellite to Inconel. The licensee refurbished multiple SOVs under this design change from 2002 to approximately 2005. In a letter from Westinghouse to Dominion (LTR-NEM-03-1050), during refurbishment of several valves in 2003, Westinghouse identified that two Stellite valves were found to have severe damage to the lower part of the override assembly. Specifically, the end of the override assembly was either partially or totally eroded off. Westinghouse noted that the severe damage was probably due to seat leakage over an extended period of time. The inspectors noted that this damage described by Westinghouse was very similar to the damage to SOV S/N K72047- 00- 0006, which failed in 2023.

The pilot SOV that failed in 2023 had been installed during an inadvertent safety injection actuation on April 17, 2005 (i.e., the tin whisker event). During this event, the PORVs cycled many times as designed to prevent the pressurizer safety valves from lifting. After this event, the licensee observed that both PORVs had excessive seat leakage.

During the associated 2005 maintenance outage following the inadvertent safety injection actuation, both PORVs were replaced with refurbished valves that had the Inconel ball and upper seat.

When the leaking SOV, serial number K72047-00-0006, was removed from service, it was placed in the fuel building in a satellite quality assurance storage area, where it remained until 2021. The valve was never refurbished with the Inconel parts and was not entered back into the warehouse s inventory tracking system.

In 2021, due to PORV leakage documented under CR1165713 and CR1180415, the decision was made to replace both the A and B PORVs during the next refueling outage. At that time, there was not a viable replacement pilot SOV to be used in support of the PORV rebuild. As a result, under PO 70384066, blocked stock pilot SOV S/N K72059- 35- 0011 and S/N K72047- 00-0006 (from satellite quality assurance storage) were sent to Westinghouse to perform further analysis on the leak tightness of the spare SOV assemblies.

Both SOVs passed the functional and seat leakage test. The valves were certified by Westinghouse, and the licensee accepted the test results.

The pilot SOV, S/N K72047- 00- 0006, was installed in the Unit 3 RCS on May 8, 2022, under WO53203161170 in the B PORV (3RCS*PCV456). The B PORV passed all required surveillance activities to support plant startup at the conclusion of 3R21 and was declared OPERABLE following successfully stroking during surveillance testing on May 18, 2022.

The licensees cause evaluation stated, The non-preferred Stellite 6 and revised Inconel 718 used the same stock-code (M2930491). Using the same stock-code prevented the ability to distinguish between preferred and non-preferred material when ordering parts.

The inspectors noted that these SOVs are ASME Class 1 pressure boundary components and are marked with a unique serial number that was traceable to its fabrication and procurement records. The inspectors also noted that while the licensee may assign a common stock-code number, each component receives a unique batch number that is also traceable to the fabrication and procurement records. Moreover, the inspectors identified that the system engineering notebook contained documentation that clearly identified the operational history for pilot SOV S/N K72047- 00-0006, which documented that the valve had not been refurbished with the Inconel parts. Other records such as procurement documents, WOs, and CRs contained sufficient information to establish that SOV S/N K72047- 00- 0006 was never overhauled. Therefore, based on the above, the inspectors determined that it was within the licensees ability to identify that the failed SOV had not been refurbished with the Inconel parts prior to its installation in the RCS on May 8, 2022.

The failure to adequately identify and control SOV S/N K72047-00- 0006, after it was removed from service in 2005, resulted in the installation and use of an incorrect component in a risk -significant, safety-related system.

Corrective Actions: The licensee replaced the failed SOV with one that contained Inconel internal parts, completed a level of effort evaluation (CA12192342), and began working on a cause evaluation (CR1257438). On May 2, 2024, the licensee submitted LER 2023- 006- 00, Millstone Power Station Unit 3, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications, and a revision to the LER on May 20, 2024. The revised LER states that the supplemental report will be sent after the licensee completes its evaluation of the event.

Corrective Action References: CR1241058, CR1242536, CR1254655, CR1257438

Performance Assessment:

Performance Deficiency: The licensee failed to meet 10 CFR Part 50, Appendix B, Criterion VIII, "Identification and Control of Materials, Parts, and Components, when it failed to prevent the installation and use of an incorrect pilot SOV (S/N K72047-00- 0006) for the B pressurizer PORV (3RCS*PCV456). Specifically, the licensee installed an SOV that contained Stellite internal parts, which were known to be susceptible to leakage when exposed to steam and contributed to the failure of the B PORV to perform its design function to open on October 20, 2023.

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. The 'B' pressurizer PORV was unable to perform its mitigating probabilistic risk assessment function to open during an initiating event.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Based on Appendix A, Exhibit 2, question 3, a detailed risk evaluation is required because the degraded condition, an inoperable pressurizer PORV, represented a loss of the probabilistic risk assessment function of one train of a multi-train technical specification system for greater than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation using Millstone Unit 3 Simplified Plant Analysis Risk (SPAR) Model, Version 8.82, with Systems Analysis Program for Hands-on Integrated Reliability Evaluation (SAPHIRE) software, Version 8.2.10.

The analyst assumed that the performance deficiency resulted in the Unit 3 pressurizer PORV, 3RCS*PCV456 ('B' train), being in a condition where it would not have been able to perform its required function of opening on demand. Based on the nature and cause of the condition and the timeline of events described above, the analyst assumed that PORV 3RCS*PCV456 was in this condition sometime from August 17, 2022, when indications revealed it began to leak, until October 20, 2023, when it failed its operability check. Since the assessed degradation that led to failure is a mechanism that gradually affects the valve during its standby time, the analyst determined the exposure period is best assessed as T/2 (429 days / 2 = 214.5 days), consistent with Risk Assessment of Operational Events, Volume I, Section 2.4.

An initial internal events risk assessment identified transients and loss of the 301A DC bus as dominant events with dominant sequences that involved failure of the reactor protection system followed by RCS overpressure (an anticipated transient without scram (ATWS)); or inadequate AFW flow followed by failure of RCS feed and bleed functions.

The analyst further reviewed the ATWS and RCS overpressure protection modeling in the SPAR model and noted that the existing default success criteria for the RCS pressure limiting safety function was a 5-out-of-5 criteria for both pressurizer PORVs and all three pressurizer safety valves to open on demand. The analyst also noted that alternative success criteria associated with this safety function can be modeled in the SPAR model if acceptable conditions exist for core reactivity feedback, total available primary-side pressure relief capacity, and available AFW capacity for the exposure period of the degraded condition

('B' PORV failed). The analyst reviewed multiple sources of design basis information, accident analyses, and operating parameters for the facility and determined that the most appropriate success criteria is represented by conducting an unfavorable exposure time analysis using WCAP- 15831. Unit 3 does not have a plant-specific unfavorable exposure time analysis, but a generic bounding analysis is incorporated into the design and licensing basis analysis and, at least, the two most recent cycle- specific core reload safety analyses. For the exposure period, the RCS pressure limiting safety function was assessed to be the opening of all three pressurizer safety valves and 1-of-2 safety PORVs. The analyst adjusted the logic structure of the existing SPAR model basic events to reflect this best available information associated with this safety function.

The analyst noted while reviewing the AFW fault tree, and during discussions with the licensee, that a postulated failure of the DC bus (i.e., 301A / 301B) event, that the associated motor-driven AFW pump could be started manually by closing the AC supply breaker locally, as described by current emergency operating procedures. The current SPAR model assumes the loss of DC control power fails the associated pump(s). Based on a review of the system response and procedure, the analyst assessed this as feasible and justified and added a human action basic event and logic structure to support functional recovery during a loss of DC control power.

The analyst determined that the degraded condition was most appropriately modeled by setting the basic event PPR-PRV-CC- 456 (PORV 456 FAILS TO OPEN ON DEMAND) to TRUE. In addition, the following SPAR model modifications were made based on plant conditions, design, and discussions with the licensee:

3RCS*MV8000B, PORV 456 upstream block valve was closed by Operations staff on August 19, 2022, in response to the leaking PORV 456 for the duration of the operating cycle. Basic event PPR-MOV-AP- BLK2 (PORV 2 Block Valve is Closed during Full Power) was set to TRUE. Since this valve was closed, it would not fail in its normally open position; basic event PPR-MOV-OO-8000B (PORV 456 Block Valve 8000B Fails to Close) was set to FALSE; also, PPR-PRV-OO-456L (PORV 456 Fails to Reclose After Passing Liquid) was set to FALSE.

RCS-POWER- HIGH was set to TRUE, to account for full power reactor operation.

RCS-PHN- MODPOOR was set to FALSE, to account for satisfactory moderator temperature coefficient.

After discussions with Idaho National Laboratory, errors were corrected (removed two house events) in the RCSPRESS (Failure to Limit RCS <3200PSI) fault tree. RCSPRESS- 4 gate (PORV Flow Paths Are Unavailable) was changed to an AND logical gate to represent a 1- of-2 PORV success criteria to account for ATWS mitigation for existing core design and plant conditions.

AFW-MDPA and AFW-MDPB fault trees were modified to add an operator manual action (EOP 35 FR-H.1, Step 4.g) to recover the motor-driven AFW pump on loss of its DC control power (LODC-301A/B). The human error probability was set to 1.2E-2 using SPAR- H (high stress diagnosis / action).

Post-processing rule was developed to replace the RCS feed and bleed operator action (HPI-XHE-XM-FAB) with an action incorporating dependency (HPI-XHE-XM-FAB1) if concurrent with manual motor-driven AFW pump power recovery in the minimum cutset (8.8E-2 versus 2E-2).

The internal events risk assessment estimated the mean increase in core damage frequency (CDF) associated with the performance deficiency to be 5.0E-7/year (5 percent 5.5E-8, percent 1.5E-6). The dominant event sequences include transients and various loss of offsite power events with inadequate AFW flow and failure of RCS feed and bleed function. A sensitivity case was performed and determined the results were insensitive to crediting of diverse and flexible coping (FLEX) strategies. External events were assessed and resulted in a seismic event risk increase contribution estimate of 2.5E-8/year and 1.1E-7/year for high winds (i.e., hurricane, straight line winds, and tornadoes).

The SPAR model does not include modeling of fire events nor does the licensee have an approved fire probabilistic risk assessment. However, the licensee is in the process of completing a fire probabilistic risk assessment and provided preliminary fire zone fire damage sequences and associated fire ignition sequences to support the analysts fire risk estimate.

The analyst requested damage sequences and ignition frequencies related to plausible fire zone scenarios that affected functions in the internal events minimum cutset sequences. This information is considered best available information and was used in conjunction with information from Millstone 3 Plant Safety Study (Section 2.5, Fire Analysis) and Individual Plant Examination, which includes the Individual Plant Examination of External Events. For each scenario a change set was developed in SAPHIRE representing a base case and conditional case (i.e., includes the 'B' PORV degraded condition), using the transient initiator (IE-TRANS) as a fire initiator surrogate. The result was multiplied by the sum of fire ignition frequencies for those fire scenarios. An incremental contribution for each scenario was calculated and summed to estimate the total fire risk. A summary of each fire zone damage state scenario and risk increase contribution follows:

'A' PORV failed, 1.8E -9 / year

'A' and 'B' motor-driven AFW failed, 1.7E-8 / year

'A' PORV and 'A' and 'B' motor -driven AFW failed, 6.5E-8 / year

'A' PORV and turbine- driven AFW failed, 2.5E-9 / year

'A' motor-driven AFW and turbine- driven AFW failed, 1.9E-10 / year

'A' PORV and 'A' motor -driven AFW failed, 1.4E-8 / year

Total fire risk increase estimate is the sum of the six scenarios: 1.0E-7 / year.

The estimate in total risk increase in CDF is the sum of the internal and external events:

5.0E-7/year + 2.5E-8/year + 1.1E-7/year + 1.0E-7/year = 7.4E- 7/year. A large early release frequency (LERF) assessment was also made using SAPHIRE and IMC 0609, Appendix H, Containment Integrity Significance Determination Process, dated March 23, 2020. This issue was determined to be of very low safety significance since SAPHIRE calculated LERF increase <1E-7/year based on zero- factor multipliers for dominant sequences.

The licensees risk significance evaluation yielded results of 6.4E-7/year (delta- CDF) and 3.7E-8/year (delta-LERF) for internal events, respectively, with similar dominant sequences as the analysts assessment.

In summary, the increase in risk associated with the performance deficiency was 7.4E-7/year and was determined to be a finding of very low safety significance (Green).

Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate, and up- to-date documentation.

Specifically, the licensee did not create and maintain complete, accurate, and up-to-date documentation for the spare PORVs and it was not clearly labeled while in storage.

Enforcement:

Violation: 10 CFR Part 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components, requires, Measures shall be established for the identification and control of materials, parts, and components, including partially fabricated assemblies. These measures shall assure that identification of the item is maintained by heat number, part number, serial number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts, and components.

Contrary to the above, on May 8, 2022, the licensees program for the identification and control of materials, parts, and components failed to prevent the installation and use of an incorrect SOV on the 'B' pressurizer PORV (3RCS*PCV456). Specifically, the licensee installed an SOV that contained Stellite internal parts, which were known to be susceptible to leakage when exposed to steam and contributed to the failure of the 'B' PORV to perform its design function to open on October 20, 2023.

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

Failure to Establish and Implement Appropriate Procedure/Instructions for Maintenance on the Unit 2 Control Room Heating and Ventilation System Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.5] - 71111.24 NCV 05000336/2024002 - 01 Operating Open/Closed Experience A finding of low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self - revealed when the licensee failed to incorporate appropriate acceptance criteria for refilling the refrigerant charge of the Unit 2 'A' CRAC compressor within work instruction and procedures.

Consequently, an incorrect amount of refrigerant was added which resulted in the inoperability of the 'A' train of the control room emergency ventilation system for longer than its technical specification allowed outage time.

Description:

On April 7, 2024, Unit 2 control room operators noticed the control room temperature was 81 degrees Fahrenheit, which was 10 degrees higher than the normal setpoint, 71 degrees Fahrenheit. The licensee discovered that the 'A' train CRAC compressor was not running, and system indications showed a low refrigerant pressure. Further investigation revealed a significant leak on the refrigerant pressure relay, which provides protection from high and low system pressure in the compressor. Based on this, the licensee declared the 'A' train of the control room emergency ventilation system inoperable and entered Technical Specification 3.7.6.1.a, a 7-day shutdown action statement, because the OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for Operations personnel during and following all credible accident conditions. The licensee swapped to the 'B' CRAC, which operated correctly and adequately maintained control room temperature.

On April 9, 2024, the pressure relay was replaced, and pressure testing of the system was completed successfully. On April 10, 2024, refrigerant was added to the system as prescribed in WO53203413663 which stated to refill in accordance with the refrigerant control checklist.

However, this did not contain any information on the correct refrigerant charge values; therefore, workers used the written guidance available in the refrigerant charge job aid to find a charge value of 200 pounds. The system was then refilled to 200 pounds of refrigerant and returned to service. The licensee exited Technical Specification 3.7.6.1.a on April 10, 2024.

On April 14, 2024, operators again noticed a temperature increase in the control room and walked down the CRAC train 'A' to find the compressor not running. The licensee re-entered Technical Specification 3.7.6.1.a and performed extensive troubleshooting to determine the cause of the compressor failure. The licensee suspected a dirty thermal protection relay contact was the most likely cause of the issue. The licensee cleaned the relay and restarted the compressor. After starting the compressor and monitoring it for an hour, the decision to restore the CRAC train 'A' to operable was made and Unit 2 exited Technical Specification 3.7.6.1.a on April 18, 2024.

The following morning on April 19, 2024, the CRAC train 'A' compressor was again found to be not working properly. The licensee consulted with the vendor who identified that on April 10, 2024, the licensee had overcharged the CRAC system with 200 pounds of Freon. As a result, the licensee decreased the amount of refrigerant within the system to 150 pounds and slowly increased it to a final value of 168 pounds. The system was then monitored and the following day the system was successfully restored to service with no further issues.

The licensee completed a level of effort cause evaluation in response to the CRs generated from this event. The licensee identified that maintenance, outage and planning, and engineering all referenced an uncontrolled document (refrigerant charge job aide) to determine the refrigerant charge value needed for the operability of the CRAC system. This uncontrolled document was the only written guidance for the licensee staff to use for refrigerant charge values when refilling the refrigerant charge.

The licensee s cause evaluation identified that in 2012, maintenance had generated CR465240 to evaluate the appropriateness of the 200- pound refrigerant charge specified by the job aide. In response to CR465240, the licensee determined that 200 pounds was incorrect. The refrigerant charge job aide was never updated with this new value or was it questioned whether it was appropriate to be used.

The licensee reported this event in LER 05000336/2024- 001-00, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications, on June 10, 2024. The inspectors reviewed this LER to verify the accuracy and completeness of the information provided and reviewed the appropriateness of the corrective actions describe in Level of Effort Evaluation CA12444430.

Corrective Actions: The licensee took action to restore the system to operable by establishing the proper refrigerant charge. The licensee entered the issue into its corrective action program, conducted a level of effort evaluation to determine causes, eliminated the use of the uncontrolled refrigerant charge job aide, and submitted LER 05000336/2024- 001- 00, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting i n a Condition Prohibited by Technical Specifications.

Corrective Action References: CR1255747, CR1257073, CR1256509

Performance Assessment:

Performance Deficiency: The failure to establish and implement adequate safety-related maintenance procedures with appropriate acceptance criteria was a performance deficiency that was within their ability to foresee and correct and should have been prevented.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the S tructures, Systems, and C omponents Performance 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. The inspectors used IMC 0612, Appendix E, example 4.M, to inform their decision making. Specifically, the failure to establish adequate maintenance procedures led to the overcharging of the CRAC train 'A' with refrigerant which caused the system to be inoperable from April 10 to 20, 2024.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. In accordance with Exhibit 3, Section D of IMC 0609, Appendix A, the inspectors determined the finding had very low safety significance (Green) because:

(1) the finding does not represent a degradation of the radiological barrier function provided for the control room and

(2) the finding does not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere.

Cross-Cutting Aspect: P.5 - Operating Experience: The organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, maintenance personnel questioned the quality of the job aid in 2012 and a CR was generated to address this concern. During work planning for repairs to the CRAC train 'A' compressor, this relevant internal operating experience was not considered which could have prevented the system from being overcharged.

Enforcement:

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

Contrary to the above, prior to April 20, 2024, Millstone Power Station maintenance procedures and instructions did not include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Specifically, the procedure and instructions used to restore the facility 1 CRAC compressor to service after maintenance did not contain a value for a proper refrigerant charge. As a result, the licensee over charged the air compressor, which rendered the 'A' train of the control room emergency ventilation system inoperable for 10 days.

The disposition of this finding closes LER 05000336/2024- 001-00, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications.

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

Failure to Perform Vibration Measurements in Accordance with American Society of Mechanical Engineers Operation and Maintenance of Nuclear Power Plants Code Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.

7] - 71111.24 Systems NCV 05000336, 05000423/2024002- 02 Documentation Open/Closed The inspectors identified a finding of very low significance ( Green) and associated NCV of 10 CFR 50.55a(f)(4)(ii) when the licensee failed to perform inservice testing on safety - related pumps in accordance with the ASME OM Code. Specifically, the licen see failed to meet subsections ISTB - 3510(c), ISTB- 3540(a), ISTB- 3540(d), and ISTB- 3300(f) when it failed to take vibration measuremen ts in appropriate locations on the pump bearing housings and did not take measurements in the same location for each test. This resulted in nonconservative data recorded during the surveillance tests.

Description:

The Millstone Units 2 and 3 inservice testing program plans commit to following the 2012 edition of the ASME OM Code, for inservice testing of safety- related pumps. The ASME OM Code requires inservice surveillance tests be performed on Class 1, 2, and 3 pumps used to shut down the reactor to the safe shutdown condition, maintain the safe shutdown condition, or mitigate the consequences of an accident. One of the test quantities measured during the inservice surveillance is pump vibration. The measured value is compared to a reference value to determine acceptable pump operation. Reference values are measured when the pump is known to be operating acceptably before they are placed into service. Vibration data is collected in the vertical, horizontal, and axial direction depending on the location on the pump. The ASME OM Code also requires that vibration measurements be taken in the same location where the reference value was obtained during each test.

As described by M3- EV- 04- 0027, Technical Evaluation for Installation and Use of Pump/Motor Vibration Monitoring Blocks, at Millstone, vibration monitoring target blocks (VMTBs) are installed on pumps and motors to assist in acquiring vibration readings. The typical VMTB is a 1 1/4- inch carbon steel cube weighing approximately 1/2 pound and is attached to the pump bearing housing using a two- part cement. The inspectors noted that the licensee uses the VMTBs to identify the vibration measurement point on each bearing.

On April 8, 2024, the inspectors observed the licensee take vibration measurements during the inservice testing of the Unit 2 'A' reactor building component cooling water pump. During this test, the inspectors identified several concerns with the collection of vibration data.

Specifically, the inspectors observed plant personnel take vibration measurements at unmarked locations on the inboard and outboard pump bearing housings because the inboard VMTB was not present and the outboard VMTB was located behind the pump shaft safety guard.

Through discussions with the licensee, t he inspectors noted that the inboard and outboard vibration acceptance criteria specified by the surveillance procedure was based on the vibration reference value, which was taken on the VMTB.

The inspectors reviewed licensee procedure CBM- 104 Vibration Data Acquisition and Overall Vibration Analysis, which provides instructions on how to acquire the inservice testing vibration data. The inspectors identified that this procedure included a note that permitted the licensee to take vibration data at an unmarked location when the VMTB was not present. The inspectors identified to the licensee that this procedure did not meet the ASME OM Code requirements, because the vibration data must be taken at a marked location to ensure the data is taken at the same location during each test.

An extent of condition walkdown was performed and over 10 safety-related pumps were identified with missing VMTBs or VMTBs were not properly located on the pump. Therefore, the inspectors questioned whether past surveillance tests on these pumps were performed in accordance with the ASME OM Code. The inspectors also questioned how the vibration data varied when taken on the VMTB versus the pump bearing housing.

The licensee performed an evaluation and determined that in most cases the on-the- pump reading was lower than the on- the- block reading, which was nonconservative because the reference values and surveillance test acceptance criteria were based on on- the-block measurements.

The licensee reviewed past vibration data for a sample of pumps that were found with missing VMTBs and determined that adequate margin to the ASME OM Code Subsection ISTB- 6200, Corrective Action, ALERT and REQUIRED ACTION ranges was maintained.

For example, on April 8, 2024, during an inservice test of the 2A reactor building closed- cooling water pump, the inspectors noted a horizontal vibration measurement of 0.0561 at the inboard pump bearing. This reading was taken from an unmarked spot on the bearing housing because the VMTB had detached. When the licensee conducted another inservice test on July 8, 2024, after reattaching the VMTB, the recorded vibration was 0.102, marking an 81.1 percent increase. Although this value was below the acceptance criterion of 0.232 and did not trigger the ALERT limit, the significant difference between the two measurements demonstrates the importance of consistently taking vibration data from the same location as the ASME OM Code requires.

Corrective Actions: The licensee entered this issue in the corrective action program and issued a standing order to ensure that VMTBs are correctly installed prior to any inservice testing.

The licensee also created a corrective action to review its vibration monitoring program and make appropriate procedure revisions to ensure compliance with the ASME OM Code.

Corrective Action References: CR1261561, CR1260096, CR1259873, CR1259349, CR1258765, CR1257100

Performance Assessment:

Performance Deficiency: The failure to perform vibration monitoring activities on safety-related pumps in accordance with Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants, of the ASME OM Code was a performance deficiency that was within the licensees ability to foresee and correct and could have been prevented.

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. The failure to perform vibration measurements in accordance with the ASME OM Code led to non- conservative vibration readings being recorded. If left uncorrected, the use of invalid or nonconservative data could impact the licensees ability to identify if those corrective actions, such as more frequent testing or declaring a pump inoperable, required by subsection ISTB-6200, Corrective Action, of the ASME OM Code would be required to ensure the availability, reliability, and capability of safety-related pumps that respond to initiating events to prevent undesirable consequences.

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

The inspectors determined the finding was of very low safety significance (Green) by using Exhibit 2, Mitigating System Screening Questions.

The finding was not a design deficiency, and it did not represent a loss of probabilistic risk assessment function because vibration monitoring is used to detect pump bearing degradation and other pump degradation prior to pump inoperability.

Cross-Cutting Aspect: H.7 - Documentation: The organization creates and maintains complete, accurate, and up- to-date documentation.

Specifically, in this case conflicting information within the implementing procedure made workers in the field implement actions contrary to code requirements.

Enforcement:

Violation: Title 10 CFR 50.55a(f)(4)(ii), requires, in part, Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval (or the optional ASME OM Code Cases listed in NRC Regulatory Guide 1.192 as incorporated by reference in paragraph (a)(3)(iii) of this section).

Millstone Power Station Unit 2 IST Program Plan - 5th Interval, Revision 2, establishes the Code of Record for the Fifth 10-Year IST Program Interval (December 2, 2018, to December 1, 2028) as the ASME OM Code, 2012 edition, as incorporated by reference in CFR 50.55a.

Millstone Power Station Unit 3 IST Program Plan - 4th Interval, Revision 3, establishes the Code of Record for the Fourth 10- Year IST Program Interval (December 2, 2018, to December 1, 2028) as the ASME OM Code, 2012 edition, as incorporated by reference in CFR 50.55a.

ISTB- 3510(c) of the 2012 ASME OM Code states The sensor location shall be established by the Owner, documented in the plant records and shall be appropriate for the parameter being measured. The same location shall be used for subsequent tests. Instruments that are position sensitive shall be either permanently mounted, or provision shall be made to duplicate their position during each test.

ISTB- 3540(d) states The measurement points shall be clearly identified on the pump to permit subsequent duplication in both location and plane.

ISTB- 3300(f) states All subsequent test results shall be compared to these initial reference values or to new reference values in accordance with paragraph ISTB-3310 or ISTB-3320, or subparagraph ISTB- 6200(c).

ISTB- 3540(a) states On centrifugal pumps, except vertical shaft pumps, measurements shall be taken in a plane approximately perpendicular to the rotating shaft in two approximately orthogonal directions on each accessible pump bearing housing. Measurements shall also be taken in the axial direction on each accessible pump thrust bearing housing.

Contrary to the above, on April 8, 2024, the licensee s inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, did not comply with the requirements of the 2012 ASME OM Code as incorporated by reference in 10 CFR 50.55a for the current 10-year inservice test program interval at Millstone Power Station, effective December 2, 2018. Specifically, for multiple safety-related pumps at Units 2 and 3, the inspectors identified the following noncompliances with the ASME OM Code, 2012 edition:

• ISTB- 3510(c): The licensee failed to use the same sensor location, as was used for the reference measurement, during subsequent inservice tests.

• ISTB- 3540(d): The licensee failed to clearly identify the measurement points on some pump bearing housings.

• ISTB- 3300(f): When vibration measurements were not taken on a VMTB, the licensee failed to establish new reference values.

• ISTB- 3540(a): The licensees use of a single VMTB to measure the horizontal, vertical, and axial vibrations was a failure to take measurements in a plane approximately perpendicular to the rotating shaft in two approximately orthogonal directions on each accessible pump bearing housing.

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

Failure to Correct a Degraded Condition on the Main Generator Output Breaker Compressed Air System Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.11] - 71152A FIN 05000423/2024002- 03 Challenge the Open/Closed Unknown A self- revealed finding (FIN) of very low safety significance (Green) was identified for failure to correct a degraded condition on the MGOB compressed air system in accordance with procedure MP - PROC- 000 - PI- AA- 200, Corrective Action.

As a result, a moisture and oil buildup occurred in the 'B' phase supply pole air channel.

This caused a ground fault on the

'B' phase of the MGOB.

The protection relaying actuation resulted in a turbine trip and a reactor trip on May 30, 2023.

Description:

The inspectors performed an in- depth review of the licensee s actions following the Unit 3 turbine generator and reactor trip on May 30, 2023 (LER 05000423/2023- 001 - 00, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault ), due to a ground fault of the 'B' phase of the MGOB. The inspectors reviewed the cause analysis and the corrective actions taken and planned. The inspectors assessed the licensee s problem identification threshold, prioritization of the issue, level of e ffort evaluation, use of operating experience, and timeliness of corrective actions. The sample was selected due to the impact on the Initiating Event cornerstone.

On May 30, 2023, while Millstone Power Station Unit 3 was in Mode 1 at 100 percent reactor power, a main generator/turbine trip occurred that resulted in a reactor trip. The main generator/turbine trip was caused by ground fault on the 'B' phase MGOB. During normal operation, plant power is provided to the plant (6.9kV) and emergency (4.16kV) buses through the two normal station service transformers (NSSTs). The NSSTs are supplied 24kV through the main generator isolated phase bus duct. Following the main generator/turbine trip, station electrical buses fast transferred from the NSSTs to the reserve station service transformers that are powered from offsite power sources.

The events leading up to the trip are as follows: An MGOB air system leak within the capacity of the compressors was identified in the Unit 3 Operations log on March 6, 2023, at 9:13 p.m.

,

and CR1221097 was submitted to address the leak. On May 13, 2023, the plant equipment operator reported MGOB local alarm 4- 5, Excessive Running Time Compressor 'B'.

The 'B' compressor was unable to maintain receiver pressure with the air leak, so the 'B' compressor was removed from service and the 'A' compressor was protected. Operations investigation determined that gross air leakage was observed at either the Priority valve or fourth stage safety valve. Due to location of components, Operations was unable to identify the source.

The MGOB air system continued to be monitored with a leakage rate that was stable at approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of runtime on the 'A' MGOB air compressor per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A low- pressure bottle air supply was made ready on May 14, 2023, that would bypass the source of the leak, but it was not installed while the compressor provided adequate makeup.

The 'B' compressor was replaced and returned to service on May 29, 2023, at approximately 9:00 a.m. The trip occurred that night on May 30, 2023, at 4:46 a.m.

The direct cause of the trip was air leaks and excessive compressor runtime, which resulted in moisture and oil accumulation in the fiberglass air supply tube of 'B' phase MGOB. The accumulated moisture caused the ground fault.

Corrective Actions: The licensee replaced the 'B' phase air supply tube, fixed air leaks in the pressure reduce cabinet, and restored the 'B' air compressor.

The licensee took interim and compensatory actions which included increasing MGOB compressor monitoring. They revised the alarm response procedure OP 3353.MGB to kick off staging backup gas resource if runtime is more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> a day. These compensatory actions are still in place.

Performance Assessment:

Performance Deficiency: The failure to take corrective actions to prevent a reactor trip in accordance with MP- PROC-000- PI-AA-200, Corrective Action, was a performance deficiency that was reasonably within the licensees ability to foresee and prevent.

Section 1.5 of MP-PROC-000- PI-AA- 200 is as follows: This procedure establishes the measures to be taken to assure that conditions adverse to quality (e.g., failures, malfunctions, deficiencies, defective material and equipment, and nonconformances) are promptly identified and corrected. MP-PROC-000- PI-AA- 200, Attachment 1, provides examples of issues requiring a condition adverse to quality condition report. Item 40 is: Planned or unplanned unit power reductions due to issues with equipment monitored by the Maintenance Rule Program. This includes the MGOB compressed air system. The licensee failed to take measures for a condition adverse to quality (i.e., Equipment within Mrule Scope) to prevent a reactor trip.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correct or mitigate the excessive run time of the MGOB compressor resulted in a ground fault and subsequent Unit 3 reactor trip.

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

This issue screened as very low safety significance (Green). Although the finding caused a reactor trip, it did not result in the loss of mitigating equipment relied upon to transition to a stable shutdown condition. Upon a loss of the NSSTs, offsite power successfully transferred to the reserve service station transformers and the units emergency diesel generators remained available.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the significant air leaks reported in March 2023 were not fixed in a timely manner, which resulted in the plant trip on May 30, 2023, due to excessive run time of the compressor. The licensee failed to understand the risk associated with the compressed air system for the MGOB. As a result, repairs were not properly prioritized to prevent the trip.

The disposition of this finding closes LER 05000423/2023- 001-00, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault.

Enforcement:

The inspectors did not identify a violation of regulatory requirements associated with this finding.

Observation: Service Water Strainer Packing Leakage Results in Corrosion of 71152S Strainer Shell The inspectors reviewed the licensees corrective action program for potential adverse trends for active Units 2 and 3 service water strainer packing leakage that might be indicative of a more significant safety issue. The inspectors observed that chronic leakage from the service water strainer packing area leakage has been ongoing for several years and recently led to degradation of the 2A, 2C, and 3B service water strainer shells. The corrosion was significant enough that in isolated locations the shell thickness was found to be less than the code minimum requirements. The licensee accepted the nonconforming conditions through issuance of operability determinations and supporting engineering technical evaluations for each strainer.

The inspectors evaluated the licensee s corrective actions and extent of condition reviews associated with the service water strainers at both Units 2 and 3. The inspectors reviewed the operability determinations and associated compensatory actions. The inspectors also reviewed the supporting structural integrity evaluations. The inspectors also reviewed completed and planned corrective actions which included increased monitoring of the strainers, more frequent packing adjustments to reduce the service water leakage, and removal of insulation and reapplication of durable coatings.

The inspectors determined that the licensees corrective action, both long-term and short-term, were appropriate to ensure that further degradation of the service water strainers at both units would be mitigated and detected prior to exceeding ASME OM Code wall thickness requirements. The inspectors determined that service water leakage from the strainer packing area represented an adverse trend but did not indicate the existence of a more significant safety issue. No more than minor violations of regulatory requirements were identified.

EXIT MEETINGS AND DEBRIEFS

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

• On July 18, 2024, the inspectors presented the integrated inspection results to Michael OConnor, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Drawings 25212-26930 Piping and Instrumentation Diagram Feedwater System, Revision 49

Sheet 2

Procedures OP 3322-003 Auxiliary Feedwater Train B Revision 008

71111.07A Calculations 24-ENG- Assessment of 'C' Reactor Plant Component Cooling Water Revision 0

04648M3 Heat Exchanger (3CCP*E1C) Tube Plugging Limit

Corrective Action CR1254532,

Documents CR1262510

71111.12 Engineering ETE-MP-2012- Underground Pipe and Tank Program Life Cycle Revision 0

Evaluations 1248 Management Plan Technical Basis

ETE-MP-2013- Buried Piping Inspection, License Renewal Aging Revision 1

1052 Management Program (MP-LR-3721/MP-LR-4721)

ETE-MP-2013- License Renewal Commitment Item 3: Buried Piping Revision 0

202 Baseline and Selective Leaching Inspections

Procedures ER-AA-BPM-101 Underground Piping and Tank Integrity Program Revision 19

71111.15 Calculations CN-TA-06-93 ATWS Analysis for the Millstone Unit 3 7% Stretch Power Revision 2

Uprate (SPU) Program

ENG-04430M2 MP2 Service Water Strainer Degraded Bottom Shell Flange Revision 0

Compliance with ASME Section VIII at Maximum Allowable

Working Pressure of 100 psi, Addenda A and B

Corrective Action CR1260811

Documents

Resulting from

Inspection

Engineering ETE-MP-2024- MPS2 Service Water Strainer L-1A Structural Integrity Revision 2

Evaluations 001 Assessment Operability Determination CA12726311

ETE-NAF-2019- Documentation of the Analytical Basis for Application of Revision 0

0004 Dominion Nuclear Safety Analysis Methods to Millstone

Power Station Unit 3

ETE-NAF-2022- MPS3 Cycle 22 Reload Safety Analysis Checklist Revision 0

0018

ETE-NAF-2023- MPS3 Cycle 23 Reload Safety Analysis Checklist Revision 0

0079

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.15 Miscellaneous Millstone Unit 3 Plant Safety Study - Fire Analysis Amendment 2;

April 2, 1984

NUSCO 171 Millstone Unit 3 Individual Plant Examination for Severe August 1990

Accident Vulnerabilities

WCAP-15831 WOG Risk-Informed ATWS Assessment and Licensing Revision 1, 2;

Implementation Process August 2007

Operability Operability Determination for the 2C Service Water Pump

Evaluations Discharge Strainer Wall Thickness Less Than Code

Minimum (CR1260581)

Procedures EOP-35 FR-H.1 Response to Loss of Secondary Heat Sink Revision 028

71111.24 Procedures SP 2611A 'A' RBCCW Pump Tests Revision 18

SP 3622.3-1 TDAFW Pump Operational Readiness and Quarterly IST Revision 27

Group B Pump Test

71152S Operability 3SWP*STR1B 'B' Service Water Pump Discharge Strainer Degradation

Evaluations

71153 Corrective Action CR1233945

Documents CR1236534

CR1247313

CR1250708

CR1260581

26