IR 05000336/2024001

Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation

ML24135A259
Person / Time
Site:	Millstone
Issue date:	05/14/2024
From:	Matt Young NRC/RGN-I/DORS
To:	Carr E Dominion Energy
References
IR 2024001
Download: ML24135A259 (1)

Text

May 14, 2024

SUBJECT:

MILLSTONE POWER STATION, UNITS 2 AND 3 - INTEGRATED INSPECTION REPORT 05000336/2024001 AND 05000423/2024001 AND APPARENT VIOLATION

Dear Eric Carr:

On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Millstone Power Station, Units 2 and 3. On May 2, 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.

Section 71111.12 of the enclosed report discusses a finding with an associated apparent violation (AV) for which the NRC has not yet reached a preliminary significance determination.

This involved an inspector identified AV of Millstone Unit 2 Operating License Condition 2.C.(11), which was associated with the failure to perform inspections of the condensate storage tank pipe trench prior to the period of extended operation to evaluate for the effects of aging.

The significance of this finding could not be determined because the condition of the pipe trench and piping within is unknown. The finding is pending a significance determination, which cannot be performed until sufficient information on the condition of condensate storage tank pipe trench structures, systems, and components is available after required inspections.

Section 71111.15 of the enclosed report discusses a finding with an associated AV for which the NRC has not yet reached a preliminary significance determination. This involved a self -revealed AV 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, associated with the licensee's failure to prevent the installation and use of an incorrect solenoid-operated valve on the 'B'

pressurizer power-operated relief valve (3RCS*PCV456).

We intend to issue our final safety significance determination and enforcement decision, in writing, within 90 days from the date of this letter. The NRCs significance determination process is designed to encourage an open dialogue between your staff and the NRC; however, neither the dialogue nor the written information you provide should affect the timeliness of our final determination. We ask that you promptly provide any relevant information that you would like us to consider in making our determination. We are currently evaluating the significance of this finding and will notify you in a separate correspondence once we have completed our preliminary significance review. You will be given an opportunity to provide additional information prior to our final significance determination unless our review concludes that the finding has very low safety significance (i.e., Green).

One Severity Level IV violation without an associated finding 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 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 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.

The inspection report describes a Severity Level IV violation which represents a compliance issue of no or relatively inappreciable potential safety or security consequences. We determined that additional follow-up using Inspection Procedure 92702, Follow-up on Traditional Enforcement Actions Including Violations, Deviations, Confirmatory Action Letters, Confirmatory Orders, and Alternative Dispute Resolution Confirmatory Orders was not warranted because the violation can be appropriately sampled using baseline inspection resources.

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 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: 05000336/2024001 and 05000423/2024001

Enterprise Identifier: I-2024-001- 0050

Licensee: Dominion Energy Nuclear Connecticut, Inc.

Facility: Millstone Power Station, Units 2 and 3

Location: Waterford, CT

Inspection Dates: January 1, 2024 to March 31, 2024

Inspectors: J. Fuller, Senior Resident Inspector E. Bousquet, Resident Inspector S. Bruneau, Project Engineer B. Dyke, Operations Engineer N. Warnek, Senior Project Engineer 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

Failure to Implement Structures Monitoring Program and Infrequently Accessed Areas Inspection Program for the Condensate Storage Tank Pipe Trench Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.5] - Work 71111.12 Systems AV 05000336/2024001- 01 Management Open The inspectors identified a finding with its safety significance yet to be determined (TBD)and associated a pparent violation (AV) of Millstone Unit 2 Operating License Condition 2.C.(11) when the licensee failed to complete certain activities described in Chapter 15, License Renewal, of the final safety analysis report (FSAR) supplement prior to July 31, 2015. Specifically, for the accessible portions of the condensate storage t ank (CST) pipe trench, the licensee did not include the area in the Infrequently Accessed Inspection Program, did not perform a baseline inspection prior to the period of extended operation, and did not perform an inspection of the area once every 5 years. As a result, the licensee had not performed an inspection of the CST pipe trench, which contains both trains of auxiliary feedwater piping, to evaluate for the effects of aging.

Unit 3 Pressurizer Power-Operated Relief Valve Failed to Open Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.11] - 71111.15 Systems AV 05000423/2024001- 02 Challenge the Open Unknown A finding with its safety significance as yet TBD and associated AV 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 exposedto steam, which contributed to the failure of the 'B' PORV to open during surveillance testing on October 20, 2023.

Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specification Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000423/2024001-03 Open/Closed An NRC-identified Severity Level IV non-cited violation (NCV) of 10 CFR 50.73(a)(2)(i)(B) was identified when the licensee failed to submit a licensee event report (LER) within 60 days from discovery of an existing, but previously unrecognized, condition prohibited by the plants technical specifications. Specifically, the licensee failed to recognize that the failure of the 'B'

pressurizer PORV to open during surveillance testing on October 20, 2023, represented a condition that existed for a time longer than permitted by the technical specifications and was reportable under 10 CFR 50.73.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000336/2023-002-00 LER 2023-002-00 for 71153 Closed Millstone, Unit 2, Failed Check Valve Resulted in an Unanalyzed Condition

PLANT STATUS

Unit 2 operated at or near rated thermal power for the entire inspection period.

Unit 3 began the inspection period at rated thermal power. On January 8, 2024, the unit was taken off line to repair a steam leak on the 'D' steam generator feedwater stop valve and to replace the 3A reactor coolant pump seal. The unit returned to rated thermal power on January 19, 2024, and remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

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

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

The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, 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 with high winds and heavy rain on January 9 and 12, 2024

71111.04 - Equipment Alignment

Partial Walkdown (IP Section 03.01) (6 Samples)

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

(1) Unit 2 accessible portions of the 'A' and 'C' service water system from the pump discharge to the strainer and piping inside the intake structure on January 3, 2024

(2) Unit 3 accessible portions of the 'C' containment recirculation system pump and associated piping outside containment, including the service water supply to the heat exchanger on January 3, 2024

(3) Unit 3 'B' quench spray system from pump suction to containment wall on February 15, 2024

(4) Unit 3 chemical and volume control system outside containment from March 12 to 13, 2024

(5) Unit 2 auxiliary feedwater piping from the CST, not including the inside of the CST pipe trench, to the suction of all three pumps on March 19, 2024

(6) Unit 2 'B' containment spray system including the pump and discharge piping and valves in the 'B' engineered safety features room from March 28 to 29, 2024

71111.05 - Fire Protection

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

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

(1) Unit 3 north containment recirculation cooler cubicle (fire area ESF-2) on January 3, 2024

(2) Unit 2 bravo emergency diesel generator cubicle (fire area A-16) on January 4, 2024

(3) Unit 3 main steam valve building north floor area (feedwater and steam generator valve areas), 41-foot elevation, during hot work (welding activities) (fire area MSV-1)on January 11, 2024

(4) Unit 3 containment (fire area RC-1) on January 11, 2024

(5) Unit 2 east piping penetration area (fire area A-10A) on March 14, 2024

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

(1) The inspectors evaluated the on-site fire brigade training and performance during an announced fire drill on February 29, 2024.

71111.06 - Flood Protection Measures

Flooding (IP Section 03.01) (2 Samples)

(1) The inspectors evaluated internal flooding mitigation protections in the Unit 3 engineered safety features building, south safety injection pump room, on March 5 and 11, 2024.

(2) The inspectors evaluated internal flooding mitigation protections in Unit 2 'B' and 'C' engineered safety features rooms on March 29, 2024

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 control room during an unplanned plant shutdown on January 8, 2024, and reactor startup activities on January 17, 2024.

(2) The inspectors observed and evaluated licensed operator performance in the control room during an emergent failure of the west direct current switchgear fan on February 22, 2024, and during a failure of the 2A service water strainer on March 11, 2024.

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

(1) The inspectors observed and evaluated Unit 3 licensed operator requalification training in the Unit 3 simulator on January 26, 2024.

(2) The inspectors observed and evaluated Unit 2 licensed operator requalification training in the Unit 2 simulator on February 27, 2024

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (3 Samples)

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

(1) Review of maintenance rule (a)(1) action plan and goals associated with the November 9, 2023, Unit 3 'A' emergency diesel generator overspeed trip (CA12237306, CA12282161, CR1243459)

(2) Unit 3 steam generator feedwater trip valve (3FWS*CTV41D) drain line weld crack led to plant shutdown on January 8, 2024 (CA12308197, CR1247719)

(3) Units 2 and 3 service water strainers including the maintenance rule (a)(1) evaluation for the 3B and 3C strainers (CR1237797, CR1238720)

Quality Control (IP Section 03.02) (1 Sample)

The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSCs remain capable of performing their intended function:

(1) Unit 3 pressurizer PORV pilot SOV which failed to actuate on October 20, 2023, during surveillance testing (work order (WO)53203161170, WO53203308167)

Aging Management (IP Section 03.03) (1 Sample)

The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:

(1) Unit 2 CST pipe trench inspections directed by the Structures Monitoring and Infrequently Accessed Areas aging management programs on March 19, 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 elevated reactor trip risk after the 'B' motor generator tripped on January 4, 2024 (CR1247563)

(2) Unit 3 elevated risk during reactor coolant system solid plant pressure control on January 9 and 10, 2024

(3) Unit 2 elevated risk associated with planned work on the 'C' reactor building component cooling water pump breaker while the 'D' circulating water pump was out of service on January 22, 2024

(4) Unit 2 elevated risk and associated mitigating actions when the reserve station service transformer was removed from service on February 9, 2024

(5) Unit 3 elevated risk and mitigating actions associated with the removal of the 'B' reserve station service transformer from service on February 27, 2024

(6) Unit 3 elevated risk and associated risk mitigating actions after the unplanned failure of the station blackout diesel generator on March 7, 2024

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (8 Samples)

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

(1) Unit 2 'A' service water strainer structural integrity evaluation for as-found external corrosion on the strainer pressure boundary components on January 2, 2024 (CR1233945, CR1247313)

(2) Unit 3 'C' containment recirculation system pump upper motor oil site glass low with active leakage present on January 4, 2024 (CR1245872, CR1247489, CR1247551)

(3) Unit 3 'A' emergency diesel generator jacket water leak during surveillance testing on January 10, 2024 (CR1244965, CR1248007)

(4) Unit 3 pressurizer safety relief valve leaking at normal operating pressure and temperature on January 17, 2024 (CR1248475, CR1248747, CR1248779)

(5) Unit 3 'A' residual heat removal system not operable due to gas void identified during performance of Surveillance Procedure (SP) 3610A.3, Residual Heat Removal System Vent and Valve Lineup Verification, on February 2, 2024 (CR1249963)

(6) Unit 3 'B' service water strainer operability determination and associated structural integrity evaluation due to external corrosion of shell results in base metal less than minimum wall thickness on February 18, 2024 (CR1250708)

(7) Unit 3 operability and structural integrity evaluation for a bent service water drain line on the 'B' engineered safety features return header piping on February 29, 2024 (CR1242065)

(8) Unit 2 extent of condition evaluation for the 'A' emergency diesel generator after the

'B' emergency diesel generator tripped due to a faulty lube oil temperature switch on March 22, 2024 (CR1254383, CR1254452)

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 3 temporary engineering change MP3-23-01138, Temporary Flexim Installation to Track MP3 'A' Reactor Coolant Pump Seal Leakage, Revision 0, associated with monitoring leakage from the 'A' reactor coolant pump seal (WO53203406889)

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated Unit 3 forced outage activities from January 8 to 18, 2024

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 (IP Section 03.01) (6 Samples)

(1) Unit 3 'C' containment recirculation cooler outlet valve (3SWP*MOV57C) after a 9-year limitorque inspection and motor operated valve static test on January 3, 2024 (WO53102826242)

(2) Unit 3 feedwater trip valve (3FWS*CTV41D) after a drain valve weld repair on January 11, 2024 (WO53203407880)

(3) Unit 3 'A' residual heat removal pump after planned maintenance on February 28, 2024 (WO53103089347)

(4) Units 2 and 3 station blackout diesel after replacement of the left bank pressure control valve and repair to the speed sensor on March 8, 2024 (WO53203411467)

(5) Unit 2 'A' service water strainer after debris caused the strainer to fail to rotate on March 11, 2024 (WO53203411944)

(6) Unit 3 'B' emergency diesel generator after planned maintenance from March 20 to 22, 2024 (WO53203387832)

Surveillance Testing (IP Section 03.01) (7 Samples)

(1) Unit 2 'A' motor driven auxiliary feedwater pump surveillances on January 4, 2024

(2) Unit 2 'B' emergency diesel generator train 'B' starting air vent valve inservice testing in accordance with SP 2624B on January 19, 2024

(3) Unit 3 'B' emergency diesel generator monthly operability tests on January 23, 2024

(4) Unit 2 'A' high-pressure safety injection pump quarterly surveillance test on January 31, 2024

(5) Unit 2 reactor protection system thermal margin / low-pressure calculator test (SP 2401J) on February 14, 2024

(6) Unit 3 'B' PORV block valve test using SP 3601F.5 and associated relay test in accordance with SP 3646A.9 on February 16, 2024

(7) Unit 3 'A' train solid state protection system operational test on March 6, 2024

Inservice Testing (IP Section 03.01) (1 Sample)

(1) Unit 2 'B' train containment spray pump inservice testing (SP 2606B) on February 20, 2024

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

(1) Unit 3 elevated unidentified leakage due to degraded 'A' reactor coolant pump seal on January 1, 2024 (CR1246269, CR1246342)

71114.06 - Drill Evaluation

Required Emergency Preparedness Drill (1 Sample)

(1) Unit 3 emergency preparedness drill with an event classification and notifications from the simulator, technical support center, and emergency operations facility on March 12,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

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

(1) Unit 2, January 1, 2023 through December 31, 2023

(2) Unit 3, January 1, 2023 through December 31, 2023

BI02: Reactor Coolant System Leak Rate (IP Section 02.11) (2 Samples)

(1) Unit 2, January 1, 2023 through December 31, 2023

(2) Unit 3, January 1, 2023 through December 31, 2023

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

Event Report (IP Section 03.02)===

The inspectors evaluated the following LERs:

(1) LER 05000336/2023-002-00, Failed Check Valve Resulted in an Unanalyzed Condition (Agencywide Documents Access and Management System Accession No. ML23243B033): The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is closed.

Reporting (IP Section 03.05) (2 Samples)

(1) The inspectors reviewed the circumstances surrounding a potential report issue involving low oil level and active leak from upper motor bearing of the Unit 3 containment recirculation system 'C' pump (CR1245872)

(2) The inspectors reviewed the circumstances surrounding a potential report issue associated with the Unit 3 pressurizer PORV pilot SOV which failed to actuate on October 20, 2023, during surveillance testing on the 'B' PORV (WO53203308167)

INSPECTION RESULTS

Failure to Implement Structures Monitoring Program and Infrequently Accessed Areas Inspection Program for the Condensate Storage Tank Pipe Trench Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.5] - Work 71111.12 Systems AV 05000336/2024001- 01 Management Open The inspectors identified a finding with its safety significance yet TBD and associated AV of Millstone Unit 2 Operating License Condition 2.C.(11) when the licensee failed to complete certain activities described in Chapter 15, License Renewal, of the FSAR supplement prior to July 31, 2015. Specifically, for the accessible portions of the CST pipe trench, the licensee did not include the area in the Infrequently Accessed Inspection Program, did not perform a baseline inspection prior to the period of extended operation, and did not perform an inspection of the area once ever y 5 years. As a result, the licensee had not performed an inspection of the CST pipe trench, which contains both trains of auxiliary feedwater piping, to evaluate for the effects of aging.

Description:

On March 19, 2024, during a walkdown of the Unit 2 auxiliary feedwater system piping from the CST to the suction of the auxiliary feedwater pumps, the inspectors noted that a section of the piping runs underground in a pipe trench with removable concrete plugs on top. This trench is referred to as the CST pipe trench.

To try and determine the current condition of the CST pipe trench and the auxiliary feedwater piping contained within, the inspectors reviewed the licensee's license renewal aging management program for structures monitoring, which is also credited for condition monitoring under the maintenance rule program.

ETE-MP-2013 - 1057, Structures Monitoring Program, Revision 1, states that the Structures Monitoring Program monitors the structures required for license renewal including the structures required by the maintenance rule which are described and evaluated in accordance with common engineering procedure C EN 104I [Condition Monitoring of Structures]. Structur al monitoring activities are intended to assess the overall integrity and condition of structures, components, support systems, and specified architectural details.

The program identifies that accessible portions of the CST pipe trench are in the scope of license renewal, the maintenance rule program, and is part of the Structures M onitoring Program. The CST pipe trench is considered an infrequently accessed area, but is considered accessible. ETE -MP-2013 -1057, Section 2.5, Inaccessible and Infrequently Accessed Areas, states, in part, For areas identified as infrequently accessed that have not otherwise been made accessible, inspected, and evaluated; a baseline inspection will be performed prior to the period of extended operation in accordance with AMP [aging management program] report MP-LR-3732/MP-LR-4732, Infrequently Accessed Areas.

Engineering evaluation of examination results will determine the need for any subsequent inspections or corrective actions.

The inspectors identified that the CST pipe trench was not included in the infrequently accessed areas aging management program (ETE-MP-2023 -1059, Infrequently Accesses Areas Inspection Program, License Renewal Aging Management Program,"

Revision 1) and did not receive a baseline inspection prior to the period of extended operation, which began in September 2015. The inspectors noted that the CST pipe trench was recently added to this program in ETE -MP-2013 -1059, Revision 2, which was issued in September 2023, but a baseline inspection was not performed.

Section 1.5, Frequency, of C EN 104I, Condition Monitoring of Structures, Revision 13, states that s tructures identified as in -scope for License Renewal will be inspected a minimum of once per 5 years.

Step 5, Inspection of Inaccessible Areas, of 14, Methodology for Inspection of Structures, states in part, infrequently accessed areas are a subset of inaccessible areas and are required by a License Renewal commitment to be inspected despite the lack of routine accessibility.

The inspectors reviewed the latest 5-year structures inspection report, which was performed in 2020 under work order ( WO) 53102841635. This work order noted that the CST pipe trench was not accessible due to the trench plugs being installed. The licensee appropriately wrote CR1138785 and created WO53203260542 to track completion of the required inspection. The inspectors identified that this work order had not been completed and was not scheduled to be performed until 2028. Upon further review, the inspectors also discovered that the structures inspection performed in 2015 (WO53102402753), prior to the period of extended operation, did not inspect the accessible portions of the CST pipe trench. Both the 2015 and 2020 work orders stated that the middle length [of the CST pipe trench] is considered an infrequently accessed area inspected as it becomes available.

Moreover, both the 2015 and 2020 inspections were performed using C EN 104I, which states in Section 5 of Attachment 14, Methodology for Inspection of Structures, that infrequently accessed areas are a subset of inaccessible areas and are required by a License Renewal commitment to be inspected despite the lack of routine accessibility.

Because the licensee had not performed the required inspections of the accessible portions of the CST pipe trench, the inspectors identified that the licensee did not properly establish and implement the license renewal programs for the Infrequently Access ed Areas and Structures Monitoring Programs prior to the period of extended operation.

Therefore, the licensee did not complete FSAR, Chapter 15, commitment items and 18.

The inspectors reviewed the licensees immediate operability evaluation in CR1254027 and noted that the auxiliary feedwater system currently meets all technical specification surveillance requirements.

Corrective Actions: The licensee entered this issue in its corrective action program as CR1254027, performed an immediate operability determination, and began working on a plan to inspect the CST pipe trench.

Corrective Action References: CR1254027

Performance Assessment:

Performance Deficiency: The licensee failed to meet License Condition 2.C.(11) for its failure to complete certain activities described in Chapter 15, License Renewal, of the FSAR supplement prior to July 31, 2015. Specifically, for the accessible portions of the condensate pipe trench, the licensee did not include the area in the Infrequently Accessed Inspection Program, did not perform a baseline inspection prior to the period of extended operation, and did not perform an inspection of the area once every 5 years.

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 the required condition monitoring inspections of the CST pipe trench, if left uncorrected, would have the potential to lead to a more significant safety concern because the aging effects on the concrete and safety-related piping supports in the CST pipe trench would not have been identified, evaluated, and corrected in the period of extended operation, which could adversely impact both trains of the auxiliary feedwater system suction piping from CST.

Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Because the licensee has not performed an inspection of the CST pipe trench, and the inspectors do not have access to directly observe current CST pipe trench conditions, the inspectors could not answer the screening questions provided by Exhibit 2 - Mitigating Systems Screening Questions. As a result, the inspectors could not screen the issue to Green and do not have enough information to complete a detailed risk evaluation. The finding is pending a significance determination, which cannot be performed until sufficient information on the condition of CST pipe trench SSCs is available after required inspections.

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee did not effectively plan or implement the inspections of the CST pipe trench.

Enforcement:

Violation: License Condition 2.C.(11) states the FSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. DENC shall complete these activities no later than July 31, 2015.

Updated Final Safety Analysis Report, Chapter 15, Section 2.1.15, Infrequently Accessed Areas Inspection Program, - Commitments states, in part:

The following program enhancements will be implemented prior to the period of extended operation:

• Program Implementation. The Infrequently Accessed Inspection Program has been established. The establishment of the Infrequently Accessed Inspection Program completes the action required for Commitment, Item 12, in Table 15.6-1.

Updated Final Safety Analysis Report, Chapter 15, Section 2.1.23, Structures Monitoring Program, - Commitments states, in part:

The following program enhancements will be implemented prior to the period of extended operation:

• Addition of Structures to the Structures Monitoring Program: The Structures Monitoring Program did not initially monitor all structures in-scope for license renewal.

The Structures Monitoring Program and the implementing procedure have been modified to include all in-scope structures. Modification of the Program and the implementing procedure completes the actions required for Commitment, Item 18, in Table 15.6-1.

Engineering Technical Evaluation, ETE-MP-2013-1057, Structures Monitoring Program, License Renewal Aging Management Program (MP-LR-3728/MP-LR-4728), Revision 1, identifies the CST pipe trench (accessible portions) as in-scope of License Renewal Program, Maintenance Rule Program, and part of the Structures Monitoring Program. Section 2.5, Inaccessible and Infrequently Accessed Areas, states that for areas identified as infrequently accessed that have not otherwise been made accessible, inspected, and evaluated; a baseline inspection will be performed prior to the period of extended operation in accordance with AMP report MP-LR-3732/MP-LR-4732, Infrequently Accessed Areas.

The ETE also states that the Structures Monitoring Program for Millstone Units 1, 2, and 3 is described in, and implemented by, Millstone Common Engineering Procedure, C EN 104I

[Condition Monitoring of Structures].

Section 1.5, Frequency, of C EN 104I, Condition Monitoring of Structures, Revision 13, states that Structures identified as in-scope for License Renewal will be inspected a minimum of once per 5 years.

Step 5, Inspection of Inaccessible Areas, of Attachment 14, Methodology for Inspection of Structures, states in part, infrequently accessed areas are a subset of inaccessible areas and are required by a License Renewal commitment to be inspected despite the lack of routine accessibility.

Contrary to the above, since July 31, 2015, the licensee failed to complete certain future activities described in the FSAR supplement prior to the period of extended operation.

Specifically, for the accessible portions of the CST pipe trench, the licensee did not include the area in the Infrequently Accessed Inspection Program, did not perform a baseline inspection prior to the period of extended operation, and did not perform an inspection of the area once every 5 years. As a result, the licensee has not performed an inspection of the CST pipe trench which contains both trains of auxiliary feedwater piping, to evaluate for the effects of aging.

Enforcement Action: This violation is being treated as an AV pending a final significance (enforcement) determination.

Unit 3 Pressurizer Power-Operated Relief Valve Failed to Open Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.11] - 71111.15 Systems AV 05000423/2024001- 02 Challenge the Open Unknown A finding with its safety significance as yet TBD and associated AV 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' pressurizer 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 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, which contributed to the failure of the B PORV to open during surveillance testing on October 20, 2023.

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, RCS 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 American Society of Mechanical Engineers ( ASME) Class 1 reactor coolant system 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 gpm), which was below the 10 gpm 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).

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 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 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 int o the warehouses 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 reactor coolant system 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 identifi ed 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 procure ment documents, work orders, and condition reports 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 identif y that the failed SOV had not been refurbished with the Inconel parts prior to its installation in the reactor coolant system on May 8, 2021.

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 root cause evaluation (CR1257438).

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 spec ification system for greater than its technical specification allowed outage time. The finding is pending a significance determination.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. The licensee did not seek to understand the operational history of a safety-related ASME Class 1 pressure boundary component that had not been tracked in the warehouse tracking system for over 15 years.

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 in correct 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 a n 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 AV pending a final significance (enforcement) determination.

Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specification Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000423/2024001 - 03 Applicable Open/Closed An NRC - identified Severity Level IV NCV of 10 CFR 50.73(a)(2)(i)(B) was identified when the licensee failed to submit an LER within 60 days from discovery of an existing, but previously unrecognized, condition prohibited by the plants technical specifications. Specifically, the licensee failed to recognize that the failure of the 'B' pressurizer PORV to open during surveil lance testing on October 20, 2023, represented a condition that existed for a time longer than permitted by the technical specifications and was reportable under 10 CFR 50.73.

Description : On August 19, 2022, the control room operators closed the 'B' PORV block valve due to excessive leakage past the PORV. The leakage was well below the technical specification limit for identified reactor coolant system leakage. Operation with one or both block valves closed is allowed by Technical Specification (TS) 3.4.4, Limiting Condition for Operation.

The licensee entered TS 3.4.

ACTION a due to the excessive seat leakage, which required that power be maintained to the block valve. This was the correct action statement to be in at the time because the licensee assumed that the valve was still able to be opened if needed.

Following the closure of the 'B' PORV block valve, tail piece temperatures downstream of the

'B' PORV continued to rise as did the level in the pressurizer relief tank. This indicated that both the block valve and the PORV were leaking. This condition continued throughout the operating cycle (427 days) and numerous condition reports were generated to document significant leakage past the block valve and the PORV.

At the beginning or 3R22, on October 20, 2023, operators were preparing for entry into low - pressure operation. Prior to low - pressure operations, SP 3601B.2 was performed, and each PORV is verified to move from closed to full open. When the 'B' PORV was tested, it failed to open. The licensee immediately declared the 'B' PORV not operable. The 'A' PORV successfully passed its surveillance test and remained Operable.

The licensee entered this issue in its corrective action program as CR1241058 and began its investigation into the cause of the failure. 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).

Millstone Unit 3 TS 3.4.4 requires that both power-operated relief valves and their associated block valves shall be OPERABLE.

With one PORV inoperable due to causes other than excessive seat leakage, TS 3.4.4 ACTION b requires, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

This action statement is similar to TS 3.4.4 ACTION a, but ACTION b requires that the power be removed from the block valve, which was not done.

The inspectors observed that although an exact date was not determined when the pilot SOV incurred damage beyond the point where the B PORV would not have functioned, the factors leading to the failure (i.e., the deterioration of SOV Stellite internals and exposure to steam leakage past the block valve) persisted for approximately 429 days. Given this, the inspectors determined that it was reasonable to infer that that the valve had been inoperable for more than three days, the TS allowed outage time, prior to the scheduled shutdown on October 19, 2023. Even though the condition was not discovered until after the allowable outage time had elapsed and the condition was rectified upon discovery, the guidance in NUREG 1022 states that an LER is required if a condition existed for a time longer than permitted by the technical specifications. Moreover, NUREG 1022 explains that the discovery date is when the event was discovered rather than when an evaluation of the event is completed. Therefore, the discovery date was October 20, 2023, and the LER was due December 19, 2023.

Corrective Actions: The licensee entered the violation in its corrective action program as CR1256886 and CR125 0419, and submitted LER 05000423/2023 - 006- 00, Pressurizer Power-Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ( ML24123A227) on May 2, 2024, 135 days late.

Corrective Action References: CR1250419, CR1256886

Performance Assessment:

The inspectors determined this violation was associated with a minor performance deficiency. Failing to submit an LER as required by 10 CFR 50.73 within the required time was a performance deficiency that was reasonably within the licensees ability to foresee and correct and should have been prevented. The inspectors reviewed the condition for Reactor Oversight Process significance and concluded there was no associated finding.

Because this violation involves the traditional enforce ment process and does not have an underlying technical violation that would be considered more than minor, a cross -cutting aspect is not assigned to this violation in accordance with IMC 0612.

Enforcement:

The Reactor Oversight Processs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: The failure to make reports to the NRC as required by 10 CFR 50.73(a)(2)(i)(B)impacted the regulatory process and was a violation of NRC requirements. The violation was processed using traditional enforcement and determined to be a Severity Level IV violation consistent with NRC's Enforcement Policy, Section 6.9.d.9, Inaccurate and Incomplete Information or Failure to Make a Required Report.

Violation: 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the holder of an operating license shall submit an LER within 60 days of discovery of the event, which includes any operation or condition which was prohibited by technical specifications.

Millstone Unit 3 TS 3.4.4 requires that both power-operated relief valves and their associated block valves shall be OPERABLE.

With one PORV inoperable due to causes other than excessive seat leakage, TS 3.4.4 ACTION b requires within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Contrary to the above, the licensee failed to submit an LER within 60 days from discovery of an existing but previously unrecognized condition prohibited by the plant technical specifications. Specifically, the licensee failed to recognize that the 'B' PORV, which failed to open during surveillance testing on October 20, 2023, represented a condition that existed for a time longer than permitted by the technical specifications, and was reportable under CFR 50.73.

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

EXIT MEETINGS AND DEBRIEFS

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

• On May 2, 2024, the inspectors presented the integrated inspection results to Michael O'Connor, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Corrective Action CR1248995

Documents CR1250178

Resulting from CR1252721

Inspection CR1252730

Drawings 25212-26912 Piping and Instrumentation Diagram Low-Pressure Safety Revision 38

Injection / Containment Recirculation, Sheet 3

212-26933 Piping and Instrumentation Diagram Service Water, Sheet 2 Revision 96

71111.06 Corrective Action CR1252600

Documents

Resulting from

Inspection

71111.12 Corrective Action CR1138785

Documents CR1245982

CR1246180

Corrective Action CR1254027

Documents

Resulting from

Inspection

Work Orders 53102402753

53102841635

203260542

71111.13 Procedures OP 3211 Solid Plant Pressure Control (ICCE) Revision 1

71111.15 Corrective Action CR0503793

Documents CR1168105

CR1206212

CR1215106

CR1241058

CR1242536

CR1257558

M3-00-3386

71111.15 Corrective Action CR1251318

Documents CR1254655

Resulting from CR1257438

Inspection

Engineering DM3-00-0472-01 Refurbished Solenoid Valve Assemblies Applicable to the 06/11/2002

Changes MP3 Solenoid Power-Operated Relief Valves (SPORVs)

Engineering ETE-MP-2023-Use-As-Is Disposition of CR1242065 for Drain Valve Revision 0

Evaluations 2024 Assembly 3SWP*V740 Bent Piping

ETE-MP-2024-MPS2 Service Water Strainer L-1A Structural Integrity Revisions 0

1001 Assessment and 1

Miscellaneous Procurement 05/05/2021

Order 70384066

Westinghouse Offer for Millstone 3 Power-Operated Relief Valve 01/17/2001

Letter LTR-NEM-Refurbishment

01-5

Westinghouse Millstone PORV Solenoid Repair 12/12/2003

Letter LTR-NEM-

03-1050

Work Orders 53203161170

203308167

203399884

M3 05 06294

M3 05 06311

71111.24 Corrective Action CR1253004

Documents

Resulting from

Inspection

Procedures SP 2610AO-1 'A' AFW Pump and Recirculation Check Valve IST, Facility 1 Revision 1

SP 3446B11 Train A Solid State Protection System Operational Test Revision 21

SP 3646A.2 Emergency Diesel Generator B Operability Tests Revision 34

Work Orders 53203389857

71151 Corrective Action CR1251049

Documents

Resulting from

Inspection

71153 Corrective Action CR1245872

Documents

Corrective Action CR1249141

Documents CR1250419

Resulting from CR1256886

Inspection

2