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Category:Letter
MONTHYEARML25022A2962025-01-23023 January 2025 Technical Specification Bases Pages ML25023A0392025-01-23023 January 2025 Senior Reactor and Reactor Operator Initial License Examinations L-25-001, Endorsement of Framatomes Requested Review Schedule for Topical Report ANP-10358P, Revision O Increased Burnup for PWRs2025-01-23023 January 2025 Endorsement of Framatomes Requested Review Schedule for Topical Report ANP-10358P, Revision O Increased Burnup for PWRs ML25014A2862025-01-21021 January 2025 Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles IR 05000336/20244032025-01-13013 January 2025 Cyber Security Inspection Report 05000336/2024403 and 05000423/2024403 (Cover Letter Only) ML24365A2032024-12-30030 December 2024 ISFSI - 10 CFR 72.30 Decommissioning Funding Plan ML24351A2422024-12-19019 December 2024 Reactor Coolant System Alloy 600 Inspection Program for License Renewal Commitment No. 15 ML24351A2412024-12-19019 December 2024 Reactor Vessel Internals Inspections Aging Management Program Submittal Related to License Renewal Commitment No. 13 ML24346A2532024-12-12012 December 2024 Correction to Implementation Date for License Amendment No. 291 Regarding Framatome Gaia Fuel IR 05000336/20243012024-12-0606 December 2024 Initial Operator Licensing Examination Report 05000336/2024301 ML24324A0442024-12-0303 December 2024 Issuance of Amendment No. 292 Regarding Adoption of Technical Specification Task Force Traveler TSTF-421 05000336/LER-2024-002, Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2024-11-26026 November 2024 Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2024-002, Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary2024-11-26026 November 2024 Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary ML24296B2352024-11-19019 November 2024 Issuance of Amendment No. 291 License Amendment to Support the Implementation of Framatome Gaia Fuel (EPID L-2023-LLA-0150) (Non-Proprietary) IR 05000245/20240012024-11-12012 November 2024 Safstor Inspection Report 05000245/2024001 ML24317A2562024-11-12012 November 2024 Core Operating Limits Report, Gycle 30 IR 05000336/20240032024-11-0707 November 2024 Integrated Inspection Report 05000336/2024003 and 05000423/2024003 and Apparent Violation and Independent Spent Fuel Storage Installation Inspection Report 07200008/2024001 ML24289A0152024-10-21021 October 2024 Review of the Fall 2023 Steam Generator Tube Inspection Report 05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024 Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary IR 05000336/20244022024-10-0808 October 2024 Security Baseline Inspection Report 05000336/2024402 and 05000423/2024402 (Cover Letter Only) ML24281A1102024-10-0707 October 2024 Requalification Program Inspection 05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status ML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24260A2192024-09-16016 September 2024 Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification ML24248A2272024-09-0404 September 2024 Operator Licensing Examination Approval ML24240A1532024-09-0303 September 2024 Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel IR 05000336/20240052024-08-29029 August 2024 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005 IR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 ML24221A2872024-08-0808 August 2024 Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage IR 05000336/20244412024-08-0606 August 2024 Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only) ML24212A0742024-08-0505 August 2024 Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification ML24211A1712024-07-25025 July 2024 Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change ML24200A1062024-07-22022 July 2024 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403 IR 05000336/20245012024-07-0101 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501 ML24180A0932024-06-28028 June 2024 Readiness for Additional Inspection: EA-23-144 IR 05000336/20240102024-06-26026 June 2024 Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010 ML24178A2422024-06-25025 June 2024 2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46 IR 05000336/20244402024-06-24024 June 2024 Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only) ML24281A2072024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 (Redacted Version) ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations ML24280A0012024-06-20020 June 2024 Update to the Final Safety Analysis Report (Redacted Version) ML24176A1782024-06-20020 June 2024 Update to the Final Safety Analysis Report ML24176A2622024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 ML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code 05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications ML24170A7832024-06-0606 June 2024 ISFSI, Virgil C. Summer, Unit 1, and ISFSI, North Anna, Units 1 & 2, and ISFSI, Surry, Units 1 & 2, and ISFSI - Submittal of Revision 36 of the Quality Assurance Topical Report ML24165A1292024-06-0505 June 2024 ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023 ML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities 2025-01-23
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000423/LER-2024-002, Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary2024-11-26026 November 2024 Door Latch Failure Resulted in Loss of Safety Function for Secondary Containment Boundary 05000336/LER-2024-002, Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2024-11-26026 November 2024 Two Main Steam Safety Valves Failed to Lift within the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024 Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary 05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-006-01, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-005, Oil Leakage from C RSS Pump Motor Challenged Meeting Its Mission Time Resulting in a Condition Prohibited by Technical Specifications2024-02-0808 February 2024 Oil Leakage from C RSS Pump Motor Challenged Meeting Its Mission Time Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-004, For Millstone Power Station, Unit 3 Regarding Reactor Coolant System Pressure Isolation Valves Operational Leakage Exceeded the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2023-12-19019 December 2023 For Millstone Power Station, Unit 3 Regarding Reactor Coolant System Pressure Isolation Valves Operational Leakage Exceeded the Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-003, RCS Temperature Detector Exceeded Time Response Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications2023-12-0808 December 2023 RCS Temperature Detector Exceeded Time Response Acceptance Criteria Resulting in a Condition Prohibited by Technical Specifications 05000423/LER-2023-002, Auxiliary Feedwater Control Valve Failure Resulting in a Condition Prohibited by Technical Specifications2023-11-30030 November 2023 Auxiliary Feedwater Control Valve Failure Resulting in a Condition Prohibited by Technical Specifications 05000336/LER-2023-002, Failed Check Valve Resulted in an Unanalyzed Condition2023-08-31031 August 2023 Failed Check Valve Resulted in an Unanalyzed Condition 05000423/LER-2023-001, Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault2023-07-27027 July 2023 Automatic Reactor Trip Due to Main Generator Output Breaker Ground Fault 05000336/LER-2023-001, For Millstone, Unit 2, Structural Integrity of a Train Service Water Header Piping Could Not Be Demonstrated Causing the Unit to Operate in a Condition Prohibited by Technical Specifications2023-07-0707 July 2023 For Millstone, Unit 2, Structural Integrity of a Train Service Water Header Piping Could Not Be Demonstrated Causing the Unit to Operate in a Condition Prohibited by Technical Specifications 05000423/LER-2022-002, Two Main Steam Safety Valves Installed in Wring Location Resulting in Failure to Lift within the Technical Specification Acceptance Criteria2022-05-31031 May 2022 Two Main Steam Safety Valves Installed in Wring Location Resulting in Failure to Lift within the Technical Specification Acceptance Criteria 05000423/LER-2022-001, Emergency Core Cooling and Reactor Plant Component Cooling Water Systems Inoperable for Time Greater than Allowed by Technical Specifications2022-03-24024 March 2022 Emergency Core Cooling and Reactor Plant Component Cooling Water Systems Inoperable for Time Greater than Allowed by Technical Specifications 05000336/LER-2022-001, Structural Integrity of Reactor Building Component Cooling Water Cracked Threaded Fitting Could Not Be Established Resulting in System Inoperable Longer than Allowed by Technical Specifications2022-03-18018 March 2022 Structural Integrity of Reactor Building Component Cooling Water Cracked Threaded Fitting Could Not Be Established Resulting in System Inoperable Longer than Allowed by Technical Specifications 05000336/LER-2021-002, Failed Check Valve Resulting in Unnalyzed and Operation Prohibited by Technical Specifications2022-01-0505 January 2022 Failed Check Valve Resulting in Unnalyzed and Operation Prohibited by Technical Specifications 05000336/LER-2021-001, Incorrectly Placed Spent Fuel Assemblies in Unit 2 Spent Fuel Pool2021-09-23023 September 2021 Incorrectly Placed Spent Fuel Assemblies in Unit 2 Spent Fuel Pool 05000423/LER-2020-005, Loss of Safety Function - Secondary Containment2020-11-19019 November 2020 Loss of Safety Function - Secondary Containment 05000423/LER-2020-002, 3, Automatic Reactor Trip Due to Main Generator Ground Fault2020-05-28028 May 2020 3, Automatic Reactor Trip Due to Main Generator Ground Fault 05000423/LER-2020-001, Technical Specification Allowed Outage Time Exceeded by Accident Monitoring Instrument2020-03-0505 March 2020 Technical Specification Allowed Outage Time Exceeded by Accident Monitoring Instrument 05000336/LER-2019-001, Manual Reactor Trip Due to Loss of the a Steam Generator Feed Pump2020-02-19019 February 2020 Manual Reactor Trip Due to Loss of the a Steam Generator Feed Pump 05000423/LER-2019-001, Regarding Plant Shutdown Required by Technical Specifications, Emergency Diesel Generator Exceeded Allowed Outage Time2020-02-13013 February 2020 Regarding Plant Shutdown Required by Technical Specifications, Emergency Diesel Generator Exceeded Allowed Outage Time 05000336/LER-2018-001, Regarding Loss of Both Trains of Control Room Emergency Ventilation Resulting in the Loss of Safety Function2018-12-19019 December 2018 Regarding Loss of Both Trains of Control Room Emergency Ventilation Resulting in the Loss of Safety Function 05000423/LER-2017-001, Regarding Loss of Safety Function - Secondary Containment2017-03-20020 March 2017 Regarding Loss of Safety Function - Secondary Containment 05000336/LER-2016-002, Regarding Manual Reactor Trip Due Loss of Two Circulating Water Pumps2016-09-20020 September 2016 Regarding Manual Reactor Trip Due Loss of Two Circulating Water Pumps 05000423/LER-2016-005, Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation2016-08-0909 August 2016 Technical Specification Required Shutdown and Manual Reactor Trip Due to Steam Generator Level Oscillation 05000423/LER-2016-004, Manual Reactor Trip Due to Low Hydrogen Gas Pressure in Main Generator2016-07-13013 July 2016 Manual Reactor Trip Due to Low Hydrogen Gas Pressure in Main Generator 05000336/LER-2016-001, Turbine Driven Auxiliary Feedwater Pump Room HELB Door Left Open2016-06-27027 June 2016 Turbine Driven Auxiliary Feedwater Pump Room HELB Door Left Open 05000423/LER-2016-003, Regarding Loss of Safety Function-Supplementary Leak Collection and Release System2016-06-0808 June 2016 Regarding Loss of Safety Function-Supplementary Leak Collection and Release System 05000423/LER-2016-002, Regarding Feedwater Isolation Signal Defeated Due to Wiring Error2016-03-23023 March 2016 Regarding Feedwater Isolation Signal Defeated Due to Wiring Error 05000423/LER-2016-001, Regarding Automatic Reactor Trip on Reactor Coolant System Low Flow Due to Loss of B Reactor Coolant Pump2016-03-16016 March 2016 Regarding Automatic Reactor Trip on Reactor Coolant System Low Flow Due to Loss of B Reactor Coolant Pump 05000336/LER-2015-003, Regarding Valid Actuation of the Reactor Protection System2016-01-0707 January 2016 Regarding Valid Actuation of the Reactor Protection System 05000336/LER-2015-002, Regarding Degraded Emergency Core Cooling System Check Valve2015-08-10010 August 2015 Regarding Degraded Emergency Core Cooling System Check Valve 05000336/LER-2015-001, Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time2015-06-16016 June 2015 Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time 05000423/LER-2015-001, Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function2015-04-20020 April 2015 Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function 05000423/LER-2014-004, Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function2015-02-0909 February 2015 Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function 05000336/LER-2014-007, Regarding Completion of Plant Shutdown Required by Technical Specifications2014-09-24024 September 2014 Regarding Completion of Plant Shutdown Required by Technical Specifications 05000336/LER-2014-006, Regarding Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power2014-07-24024 July 2014 Regarding Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power 05000336/LER-2014-005, Regarding Train a Containment Spray Inoperable Due to Gas Voids2014-07-16016 July 2014 Regarding Train a Containment Spray Inoperable Due to Gas Voids 05000423/LER-2014-003, Regarding Turbine Driven Auxiliary Feedwater Pump Operability Impacted by Incorrect Bearing2014-06-30030 June 2014 Regarding Turbine Driven Auxiliary Feedwater Pump Operability Impacted by Incorrect Bearing 05000336/LER-2014-004, Regarding Foreign Material Found in a Motor Lead Rendered a Motor Driven Auxiliary Feedwater Pump Inoperable2014-06-0909 June 2014 Regarding Foreign Material Found in a Motor Lead Rendered a Motor Driven Auxiliary Feedwater Pump Inoperable 05000336/LER-2014-003, Regarding Loss of Safety Function Due to Inoperable Enclosure Building2014-06-0202 June 2014 Regarding Loss of Safety Function Due to Inoperable Enclosure Building 05000336/LER-2014-002, 2 Regarding DC Circuit Hot Shorts2014-05-0505 May 2014 2 Regarding DC Circuit Hot Shorts 05000423/LER-2014-002, Regarding DC Circuit Hot Shorts2014-05-0505 May 2014 Regarding DC Circuit Hot Shorts 05000336/LER-2014-001, Regarding Completion of Plant Shutdown Required by Technical Specifications2014-03-31031 March 2014 Regarding Completion of Plant Shutdown Required by Technical Specifications 05000423/LER-2014-001, Limiting Condition for Operation Exceeded Upon Approval of Enforcement Discretion2014-03-17017 March 2014 Limiting Condition for Operation Exceeded Upon Approval of Enforcement Discretion 05000423/LER-2013-009, Regarding Secondary Containment Boundary Breach Could Have Prevented Safety Function2014-01-15015 January 2014 Regarding Secondary Containment Boundary Breach Could Have Prevented Safety Function 05000336/LER-2013-004, Unit 2 Regarding Reactor Trip While Backwashing D Waterbox2013-12-19019 December 2013 Unit 2 Regarding Reactor Trip While Backwashing D Waterbox 2024-09-26
[Table view] |
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~Dominion Dominion Nuclear Connecticut, Inc.
Rope Ferry Rd., Waterford, CT 06385 Mailing Address: P.O. Box 128 Warerford, CT 06385 domecom Afh10 20l15 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.
MPS Lic/LES Docket No.
License No.15-403 R0 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION UNIT 2 LICENSEE EVENT REPORT 201 5-002-00 DEGRADED EMERGENCY CORE COOLING SYSTEM CHECK VALVE This letter forwards Licensee Event Report (LER) 2015-002-00 documenting an event at Millstone Power Station Unit 2 on June 11, 2015. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(v)(C). A supplemental LER is planned to be submitted by December 11,2015.
If you have any questions or require additional information, please contact Mr. Thomas G.
Cleary at (860) 447-1791 x3232.
Sincerely, Plant Manager - Millstone Attachments: I Commitments made in this letter: None f695-
Serial No.15-403 Docket No. 50-336 Licensee Event Report 2015-002-00 Page 2 of 2 cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman NRC Project Manager Millstone Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C-2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station
Serial No.15-403 Docket No. 50-336 Licensee Event Report 2015-002-00 ATTACHMENT LICENSEE EVENT REPORT 2015-002-00 DEGRADED EMERGENCY CORE COOLING SYSTEM CHECK VALVE MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
'NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 01/31/2017
,(02-2014)
- ,.,.,"",oEstimated burden per reeponse to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
,,-*'t*_.**-Reported lessons learned are incorporated into the licensing process and fed back to industry.
,~,
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections
°... LICENSEE EVIENT REPORT (LER)
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by itmte-mail to lnfocollects.Resource @nrc.gov, and to the Desk Ottficer, Ottice of Information and (See Page 2 for required number of RegulatoryAffairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid 0MB digit.S/Ch aracters fo uac lock) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
[3. PAGE Millstone Power Station Unit 2 05000336J 1 OF 5
- 4. TITLE Degraded Emergency Core Cooling System Check Valve
- 5. EVENT DATE
- 6. LER NUMBER
[
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MOT A EA EUNIA E((FACILITY NAME DOCKET NUMBER MOT A ER YEAR SEUMENILR EVO MONTH DAY YEAR 05000 jFACILITY NAME DOCKET NUMBER 06 11 2015 2015 -
002
- - 00 [08 10 2015 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[]
20.2201(b)
LII 20.2203(a)(3)(ii)
[]
50.73(a)(2)(ii)(C)
[]
50.73(a)(2)(viii)(
D] 20.2203(a)(1)
LI 20.2203(a)(4)
LI 50.73(a)(2)(ii)(B)
LI 50.73(a)(2)(viii)(B)
L 20.2203(a)(2)(i)
[]
50.36(0)(1)(i)(A)
LI 50.73(a)(2)(iii)
LI 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL LI 20.2203(a)(2)(ii)
LI 50.36(c)(1)(ii)(A)
LI 50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x)
LI 20.2203(a)(2)(iii)
LI 50.36(c)(2)
LI 50.73(a)(2)(v)(A)
LI 73.71(a)(4)
LI 20.2203(a)(2)(fv)
LI 50.46(a)(3)(ii)
LI 50.73(a)(2)(v)(B)
LI 73.71(a)(5) 100 LI 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
Z] 50.73(a)(2)(v)(C)
LI OTHER
,Lf-I 20.2203(a)(2)(vi)
,LF"I 50.73(a)(2)(i)(B)
L 1-I 50.73(a)(2)(v)(D) specify in Abstract below or in
1. EVENT DESCRIPTION
On June 11, 2015, while Millstone Power Station Unit 2 (MPS2) was in MODE 1 operating at 100 percent power, Engineering identified that due to a degraded check valve, the post-accident radioactivity release rates assumed in the FSAR could be affected. While performing 'B' High Pressure Safety Injection (HPSI) pump in-service testing, the measured flow was lower than expected. Because both trains of HPSI, Low Pressure Safety Injection (LPSI) pumps, and Containment Spray (CS) pumps share a common minimum flow recirculation line back to the Refueling Water Storage Tank (RWST), back-leakage through one of the idle pumps recirculation check valves was postulated as the cause of the observed drop in flow. Troubleshooting was performed, and it was determined that a degraded minimum flow check valve associated with the 'A' CS pump (2-CS-6A) was back-leaking into the "A" train suction line and flowing to the RWST through 2-CS-14A (RWST suction check valve) and normally open 2-CS-13.IA (RWST suction line isolation).
The associated minimum flow isolation valve for 2-CS-6A was closed to eliminate the path and the valve was repaired. 2-CS-14A permitted back flow because the conditions during the test provided insufficient back flow to adequately seat it.
Subsequently, engineering evaluation determined that during a loss of power to one facility or train of the Emergency Core Cooling System (ECCS), this back-leakage flow path could result in more leakage to the RWST than had been previously considered in the accident analysis. This leakage has the potential to adversely affect calculated post-Loss of Coolant Accident recirculation phase radioactivity release rates under some postulated scenarios. Specifically, without power on the affected (i.e.: back-leaking) train, the CS discharge flowpath would not automatically open resulting in leakage to the RWST.
Prior to this discovery, the last time that the 'A' containment spray pump was run was on April 15, 2015, which would have opened 2-CS-6A. On this basis, it was concluded that this condition existed from April 15, 2015 to June 11, 2015. During this period, the 'A' Emergency Diesel Generator (EDG) was inoperable 4 times for a total of approximately 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. However, all 4 times were for surveillances. At no time during this period was any maintenance done on the 'A' EDG. It is judged that, between accident initiation and commencement of sump recirculation (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with only one train available), the EDG could have been made available.
This event is being reported as an event or condition that could have prevented fulfillment of a safety function to control the release of radioactive material under 10 CFR 50.73(a)(2)(v)(C). Note: The initial non-emergency report (#51149) of this issue on June 11, 2015 was subsequently updated on July 10, 2015. An additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.
BACKGROUND The MPS2 Containment Spray (CS) system functions as an engineered safety feature to limit containment pressure and temperature after a loss-of-coolant accident (LOCA) and'Main Steam Line Break (MSLB) accident, and thereby reduce the potential for leakage of airborne radioactivity to the outside environment. A minimum flow recirculation line is included in the design for recirculating water from the outlet of the pump to the RWST. With the CS system operating during a design basis accident, a small portion of the CS pump discharge flow recirculates to the RWST during the injection
phase; however, the recirculation line is isolated from the RWST when transferring to sump recirculation. All seven ECCSICS pumps (2 LPSl, 2 CS, and 3 H PSI) have minimum flow recirculation lines that tie into one common header to the RWST. Exhibit A attached to this LER is a sketch depicting the configuration.
Evaluation of this failure identified that during a Small Break Loss of Coolant Accident (SBLOCA),
concurrent with a loss of power to the "A" train associated with the leaking 2-CS-6A, the operating (opposite) train HPSI pump would pressurize the common recirculation header during the recirculation phase of the accident as was described above. In the operator response to this postulated event when the 2-CS-6A leaks to the RWST, suction 2-CS-I13.1A is manually isolated terminating the release to the RWST. However, it was additionally identified that the "A" train suction header would pressurize because of the continued back leakage and none of the "A" train pumps flowing due to loss of power and no open flow path (to containment spray or RWST). This pressurization would continue, exceeding the design pressure of the suction header (60 psig) ultimately reaching 500 psig, the lift setpoint for the "A" train shutdown cooling heat exchanger relief valve downstream of the "A" CS pump and the relief valve on the common LPSl discharge header.
These relief valves discharge to the EDST (Emergency Drain Sump Tank) which is located and vented outside the filtered ventilation boundary. Upon emptying of the RWST and initiation of sump recirculation, the described back-leakage flow path would contain sump fluid. This additional unidentified release path from the original design of the plant was reported to the NRC as a follow-up notification on July 10, 2015.
For this condition (2-CS-6A back-leakage) to result in significant radiological consequences, the following accident sequence must occur:
2-CS-6A back-leakage SBLOCA Loss of Offsite Power (LOOP)
Loss of the "A" train of onsite electrical power SBLOCA break size that results in RCS re-pressurization and sustained RCS pressures above 500 psig, such that HPSl would continue to pressurize the recirculation header above the relief valve setpoints Fuel damage While the discussion above relates to 2-CS-6A and the "A" train, back-leakage from any of the minimum flow check valves and a concurrent loss of power to that train would have a similar outcome.
2. CAUSE
The cause of the event was a failed-open check valve. Further analysis indicates a leaking check valve could also cause the problem described in the Assessment of Safety Consequences.
Leakage through these check valves was not considered in the FSAR.
3. ASSESSMENT OF SAFETY CONSEQUENCES
A preliminary assessment reveals that if a recirculation line check valve fails open or leaks, the following conditions could occur:
- 1. An additional recirculation fiow path exists for the running pumps back to the non-running pumps' respective suction lines via the common minimum flow recirculation line.
- 2. During a loss of power to one facility or train of the ECCS/CS systems, more leakage to the RWST than had been previously analyzed could result.
- 3. Under certain small break LOCA conditions, ECCS/CS non-operating train suction piping could be pressurized significantly above its design pressure, and relief valves could lift, sending sump recirculation water to the Equipment Drains Sump Tank (EDST), and result in an unfiltered release of radioactivity, and associated reduction in sump inventory.
- 4. Low pressure piping portions of ECCS systems could be caused to yield for the failed open check valve case, but not for the leaking check valve case, although would be unlikely to rupture.
This scenario must be evaluated using modified methods. Millstone typically evaluates design basis radiological consequences based on a LBLOCA and associated fuel damage and a SBLOCA model does not exist. Therefore, a preliminary sensitivity model has been developed using best-estimate assumptions (i.e., a gap release source term, 1% iodine partitioning in the EDST, and measured control room filter efficiency and unfiltered leakage) for application in the small break LOCA scenario. Preliminary results show that control room operator dose would remain within the regulatory limits of 10 CFR 50.67 if the release from 2-CS-6A is terminated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and offsite doses would also remain within the regulatory limits of 10 CFR 50.67 (25 REM TEDE offsite dose and 5 REM TEDE Control Room dose) under these assumptions.
The evaluation of the dose consequences is continuing for this event, and a supplement to the LER is planned once the evaluation is completed.
4. CORRECTIVE ACTION
The failed check valve 2-CS-6A has been repaired. Additional corrective actions are being taken in accordance with the station's corrective action program.
5. PREVIOUS OCCURRENCES
None
- 6. Energy Industry Identification System (EllS) codes:
Pump-P
- Safety Injection - SIU.S. NUCLEAR REGULATORY COMMISSION (02-2014)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME [2. DOCKET
- 6. LER NUMBER j
- 3. PAGE ISEQUENTIAL REV IYEAR NUMBER NO.
Millstone Power Station Unit 2 05000336 5
OF 5
2015 oo 00oo0
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05000336/LER-2015-001, Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time | Regarding Historical Oil Leakage from C RBCCW Pump Bearings Challenged Meeting Mission Time | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000423/LER-2015-001, Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function | Regarding Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000336/LER-2015-002, Regarding Degraded Emergency Core Cooling System Check Valve | Regarding Degraded Emergency Core Cooling System Check Valve | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(viii)(B) | 05000336/LER-2015-003, Regarding Valid Actuation of the Reactor Protection System | Regarding Valid Actuation of the Reactor Protection System | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) |
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