08-09-2017 | On February 11, 2017, Unit 2 was in Mode 5 for a planned refueling outage. While attempting to fill and vent the Unit 2 High Pressure Core Spray ( HPCS) system, no flow was observed from the drywell vent valves or downstream of the HPCS injection valve. The HPCS system was already inoperable to support scheduled surveillances performed on February 8, 2017 in which the HPCS injection isolation valve had been cycled five times satisfactorily. Troubleshooting determined the cause of the valve malfunction was due to stem-disc separation. The valve internal components were replaced prior to restart of the unit from the refueling outage. The root cause of the valve failure was insufficient capacity of the shrink-fit stem collar, combined with multiple high-load cycles, which resulted in loosening and eventual shear failure of the wedge pin and threads.
This component failure is reported in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the HPCS system, a single train safety system, from performing its design function if the valve failure occurred during an actual demand. This component failure is also reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS) 3.5.1 "ECCS - Operating," since the HPCS system could have been 1 inoperable for greater than the TS 3.5.1, Required Action B.2, Completion Time of 14 days to restore HPCS system to operable status. There were minimal safety consequences associated with the condition since HPCS was not required to be operable at the time of the failure, and other required emergency safety systems remained operable. There were no actual demands for Unit 2 LHPCS, other ECCS systems, or the reactor core isolation cooling (RCIC) system during this period.
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Category:Letter
MONTHYEARIR 05000373/20230042024-01-24024 January 2024 County Station - Integrated Inspection Report 05000373/2023004 and 05000374/2023004 ML24024A1332024-01-24024 January 2024 Confirmation of Initial License Examination IR 05000373/20230122024-01-18018 January 2024 County Station - Biennial Problem Identification and Resolution Inspection Report 05000373/2023012 and 05000374/2023012 ML23354A2902024-01-0505 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0028 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23360A6082023-12-27027 December 2023 County Station Request for Information for NRC Commercial Grade Dedication Inspection: Inspection Report 05000373/2024010 and 05000374/2024010 ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums RS-23-120, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information IR 05000373/20230032023-11-0909 November 2023 County Station Integrated Inspection Report 05000373/2023003, 05000374/2023003, and 07200070/2023001 ML23286A2602023-11-0808 November 2023 Issuance of Amendment Nos. 260 and 245 to Renewed Facility Operating Licenses Relocation of Pressure and Temperature Limit Curves to the Pressure Temperature Report IR 05000373/20234012023-11-0707 November 2023 County Station Security Baseline Inspection Report 05000373/2023401 and 05000374/2023401 RS-23-103, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-13013 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans IR 05000374/20230102023-10-11011 October 2023 NRC Inspection Report 05000374/2023010 IR 05000373/20233012023-09-15015 September 2023 Errata to NRC Initial License Examination Report 05000373/2023301; 05000374/2023301 RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor ML23212A9012023-08-0303 August 2023 Regulatory Audit Report to Support the Review of the Amendments to Relocation of the Pressure Temperature Limit Curves to the Pressure and Temperature Limits Report IR 05000373/20230022023-08-0101 August 2023 County Station - Integrated Inspection Report 05000373/2023002 and 05000374/2023002 ML23208A3182023-07-27027 July 2023 Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and RFI RS-23-084, Response to Request for Additional Information Regarding the Application to Revise Design Basis to Allow Use of Plastic Section Properties in Lower Downcomer Braces Analysis2023-07-24024 July 2023 Response to Request for Additional Information Regarding the Application to Revise Design Basis to Allow Use of Plastic Section Properties in Lower Downcomer Braces Analysis ML23192A5272023-07-12012 July 2023 NRC Initial License Examination Report 05000373/2023301; 05000374/2023301 ML23186A2062023-07-0606 July 2023 Information Request for a NRC Post-Approval Site Inspection for License Renewal 05000374/2023010 ML23181A1502023-06-30030 June 2023 Combined Response to Request for Additional Information and Supplemental Information in Support of LAR to Relocate Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III IR 05000373/20230112023-06-26026 June 2023 County Station - Quadrennial Fire Protection Team Inspection Report 05000373/2023011 and 05000374/2023011 RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23167A0352023-06-16016 June 2023 Registration of Use of Cask to Store Spent Fuel ML23171A9562023-06-12012 June 2023 Post Exam Ltr RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling IR 05000373/20230012023-05-10010 May 2023 County Station - Integrated Inspection Report 05000373/2023001 and 05000374/2023001 ML23123A2202023-05-0303 May 2023 Relief Request I4R-14 for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection ML23118A3472023-05-0101 May 2023 County, 1 & 2; Nine Mile Point, 2; and Quad Cities, 1 & 2 - Correction of Amendment No. 193 Adoption of TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration EPID L-2022-LLA-0143 ML23121A1612023-05-0101 May 2023 Information Meeting (Open House) with a Question-and-Answer Session to Discuss NRC 2022 End-of-Cycle Plant Performance Assessment of LaSalle County Station, Units 1 and 2 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI ML23081A0382023-04-25025 April 2023 County, 1 & 2; Nine Mile Point, 2; and Quad Cities, 1 & 2 - Issuance of Amendments to Adopt TSTF-306, Rev. 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration ML23110A3202023-04-21021 April 2023 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection ML23107A1822023-04-18018 April 2023 Operator Licensing Examination Approval Lasalle County Station, May 2023 RS-23-049, Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-23023 March 2023 Constellation Energy Generation, LLC, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations ML23073A2182023-03-22022 March 2023 Issuance of Amendment No. 258 to Renewed Facility Operating Licenses Exigent Amendment to Revise Design Basis Related to Seismic Requirements RS-23-048, Exigent Amendment Request to Revise Design Basis Related to Seismic Requirements2023-03-0707 March 2023 Exigent Amendment Request to Revise Design Basis Related to Seismic Requirements ML23061A1632023-03-0303 March 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch 3 IR 05000373/20220062023-03-0101 March 2023 Annual Assessment Letter for LaSalle County Station, Units 1 and 2 (Report 05000373/2022006 and 05000374/2022006) RS-23-045, Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.2032023-02-28028 February 2023 Constellation Energy Generation, LLC Submittal of Fitness for Duty Performance Data Reports for 2022 Per 10 CFR 26.717(c) & 10 CFR 26.203 RS-23-037, Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors for LaSalle County Station2023-02-22022 February 2023 Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors for LaSalle County Station ML23024A1372023-02-0303 February 2023 Request for Withholding Information from Public Disclosure for LaSalle County Station, Unit Nos. 1 and 2 ML23031A1972023-01-31031 January 2023 Notification of NRC Fire Protection Team Inspection Request for Information, Inspection Report Nos. 05000373/2023011 and 05000374/2023011 IR 05000373/20220042023-01-31031 January 2023 County Station - Integrated Inspection Report 05000373/2022004 and 05000374/2022004 RS-23-003, Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-102023-01-31031 January 2023 Constellation Energy Generation, LLC, Summary of Changes to Quality Assurance Topical Report, NO-AA-10, and Decommissioning Quality Assurance Program, NO-DC-10 2024-01-05
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000373/LER-2017-0072017-08-18018 August 2017 Low Pressure Core Spray System Inoperable due to Loss of Cooling, LER 17-007-00 for LaSalle County Station, Unit 1, Regarding Low Pressure Core Spray System Inoperable due to Loss of Cooling 05000374/LER-2017-0032017-08-0909 August 2017 High Pressure Core Spray System Inoperable due to Injection Valve Stem-Disc Separation, LER 17-003-01 for LaSalle County Station, Unit 2 Regarding High Pressure Core Spray System Inoperable due to Injection Valve Stem-Disc Separation 05000374/LER-2017-0042017-07-14014 July 2017 Two Main Steam Safety Relief Valves Failed Inservice Lift Inspection Pressure Test, LER 17-004-01 for LaSalle, Unit 2, Regarding Two Main Steam Safety Relief Valves Failed Inservice Lift Inspection Pressure Test 05000373/LER-2017-0062017-07-14014 July 2017 Low Pressure Core Spray Inoperable due to Minimum Flow Valve Failure in Closed Position, LER 17-006-00 for LaSalle, Unit 1, Regarding Low Pressure Core Spray Inoperable due to Minimum Flow Valve Failure in Closed Position 05000373/LER-2017-0052017-04-18018 April 2017 Manual Reactor Scram Resulting From Feedwater Regulating Valve Failure Causing High Reactor Water Level, LER 17-005-00 for LaSalle County, Unit 1, Regarding Manual Reactor Scram Resulting From Feedwater Regulating Valve Failure Causing High Reactor Water Level 05000373/LER-2017-0042017-04-17017 April 2017 Secondary Containment Inoperable Due to Interlock Doors Open, LER 17-004-00 for LaSalle County, Unit 1, Regarding Secondary Containment Inoperable Due to Interlock Doors Open 05000373/LER-2017-0032017-04-14014 April 2017 Automatic Reactor Scram due to Main Generator Trip on Differential Current During Back-Feed Operations, LER 17-003-00 for LaSalle County Station, Unit 1, Regarding Automatic Reactor Scram due to Main Generator Trip on Differential Current During Back-Feed Operations 05000374/LER-2017-0022017-03-30030 March 2017 High Pressure Core Spray System Declared Inoperable due to Cooling Water Strainer Backwash Valve Stem-Disc Separation, LER 17-002-00 for LaSalle, Unit 2, Regarding High Pressure Core Spray System Declared Inoperable due to Cooling Water Strainer Backwash Valve Stem-Disc Separation 05000374/LER-2017-0012017-03-24024 March 2017 Manual Reactor Scram due to Turbine-Generator Run-Back Caused by Stem-Disc Separation in Stator Water Cooling Heat Exchanger Inlet Valve, LER 17-001-00 for LaSalle County, Unit 2 Regarding Manual Reactor Scram due to Turbine-Generator Run-Back Caused by Stem-Disc Separation in Stator Water Cooling Heat Exchanger Inlet Valve 05000373/LER-2017-0012017-02-0808 February 2017 Reactor Core Isolation Cooling System Inoperable Longer than Allowed by the Technical Specifications due to Low Suction Pressure Trips, LER 17-001-00 for LaSalle County, Unit 1, Regarding Reactor Core Isolation Cooling System Inoperable Longer than Allowed by the Technical Specifications due to Low Suction Pressure Trips 05000374/LER-2016-0012016-04-18018 April 2016 Secondary Containment Inoperable Due to Interlock Doors Open, LER 16-001-00 for LaSalle County, Unit 2, Regarding Secondary Containment Inoperable Due to Door Interlock Doors Open 05000373/LER-2016-0012016-04-11011 April 2016 Secondary Containment Inoperable due to Reactor Building Ventilation Damper Failure, LER 16-001-00 for LaSalle County, Unit 1 & 2, Regarding Secondary Containment Inoperable Due to Reactor Building Ventilation Damper Failure 2017-08-09
[Table view] |
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
PLANT AND SYSTEM IDENTIFICATION
LaSalle County Station Unit 2 is a General Electric Boiling Water Reactor with 3546 Megawatts Thermal Rated Core Power.
The affected system was the Division 3 High Pressure Core Spray (HPCS) system, one of the stand-by emergency core cooling systems (ECCS) credited for emergency injection into the reactor pressure vessel (RPV). The HPCS system is designed to provide sufficient cooling to the reactor core to prevent excessive fuel cladding temperatures following any break in the nuclear system piping. The affected component was the motor operated HPCS injection isolation valve (2E22-F004). This valve is normally closed and automatically opens following a HPCS injection signal to allow injection to the RPV from the HPCS pump.
CONDITION PRIOR TO EVENT
Unit(s): 2 Date: February 11, 2017 Time: 1200 CST Reactor Mode(s): 5 Mode(s) Name: Refueling Power Level: 0 percent
DESCRIPTION
During the Unit 2 refueling outage, with the reactor in Mode 5, the Unit 2 HPCS system was declared inoperable on February 8, 2017 to support performance of the HPCS high pressure water leak rate test and stem lubrication and rotation check of the HPCS injection isolation valve 2E22-F004. On February 11, 2017, while attempting to fill and vent the HPCS system, problems arose when finishing the drywell portion of the fill. Full flow water was observed from the high point vents (valves 2E22-F349 / 2E22- F350) with the water leg pump and cycled condensate systems lined up for fill. With the HPCS injection isolation valve 2E22-F004 opened from the main control room, and the HPCS check valve 2E22-F005 pinned open, no air or water was observed through the vents in the drywell. The check valve 2E22-F005 was cycled to verify proper operation, and the HPCS injection isolation valve 2E22-F004 was also cycled to verify the valve was open. However, again no air or water was observed from the drywell vents. Additional trouble shooting was performed that determined there was no flow downstream with the valve open.
Prior to this fill and vent sequence, the HPCS system had been taken out of service for leak rate testing and then drained for relief valve work. The leak rate tests (which involved cycling the 2E22-F004 valve open and closed) all passed satisfactory. Upon completion of those tests, the system was drained from the drywell down to the pump suction. System parameters observed during the leak rate tests provided firm evidence that the HPCS injection isolation valve satisfactorily cycled as designed.
Therefore, it was concluded that the HPCS injection isolation valve 2E22-F004 failure occurred sometime after the successful leak rate tests, and most likely during the fill and vent sequence. The Unit 2 HPCS injection valve had been successfully cycled open and closed five times during previous surveillances required by Technical Specifications (TS) during the refueling outage.
CAUSE
Upon investigation, troubleshooting determined the cause of the valve malfunction was due to stem-disc separation. The valve stem threads and wedge pin were found to be damaged, causing separation from the valve disc. There have been industry issues documented in a Flowserve 10 CFR Part 21 Notification concerning the quality of the wedge pin connections of Anchor Darling double disc gate valves. Topical Report BWROG-TP-13-006 documents instances where this type of valve has stem to upper wedge threaded connection failures caused by the valve sterns not being properly torqued into the upper wedge. This operating experience suggests that vendor quality was a causal factor for the component failure.
Valve disassembly and inspection revealed the wedge pin to be sheared, and the valve stern threads damaged, causing the stem to separate from the valve disc. Anchor Darling double disc gate valves are unique in design as the disc assembly consists of dual floating discs with a two piece wedging mechanism between them. The valve stem is threaded and torqued into the upper wedge. A hole is drilled through the stem for the wedge pin to hold the disc retainers in place. There have been instances documented within BWROG-TP-13-006 that state the cause of stem disc separation was the stem was not properly torqued into the upper wedge. BWROG-TP-13-006 states that there is no non-intrusive test or inspection method to determine if the stems were adequately torqued into the upper kedge prior to pin installation. Flowserve recommends that all critical Anchor Darling 1 double disc gate valves with threaded stem to upper wedge connections and actuators that produce torque be evaluated for ; comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
U.S. NUCLEAR REGULATORY COMMISSION potential wedge pin failure. At the time of the failure, the station was following vendor and industry guidance for inspection of the valve related to the Flowserve Part 21 Notification.
The root cause of the HPCS injection isolation valve 2E22-F004 stem-to-wedge connection failure was insufficient capacity of the shrink-fit stem collar, combined with multiple high load closing cycles (with both axial thrust and torque components), resulting in loosening and eventual shear failure of the wedge pin and threads. A contributing cause was insufficient preload and insufficient capacity of the stem collar and wedge pin assembly. In particular, the collar axial load capacity was 50 to 60 percent of the normal applied loads, allowing collar slippage along the stem to occur.
REPORTABILITY AND SAFETY ANALYSIS
This component failure is reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS due to the failure to complete TS 3.5.1, "ECCS — Operating," Required Action B.2, to restore HPCS system to operable status within the specified Completion Time of 14 days. This component failure is also reported in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the HPCS system, a single train safety system, from performing its design function.
There were minimal safety consequences associated with the condition since HPCS was not required to be operable at the time of discovery of the condition, and other required emergency safety systems remained operable. The HPCS injection isolation valve 2E22-F004 was cycled open and closed satisfactorily during previous flow surveillances. There were no actual demands for Unit 2 HPCS, other ECCS systems, or the reactor core isolation cooling (RCIC) system during this period.
This condition was determined to not be a safety system functional failure (SSFF) as defined in accordance with NEI 99-02, Regulatory Assessment Performance Indicator Guideline. The failure analysis demonstrated that valve failure occurred during the refueling outage when HPCS was not required to be operable. Therefore, the HPCS system was fully functional for the past operating cycle. During the past operating cycle, the valve was capable of performing its design function, which is four required cycles.
CORRECTIVE ACTIONS
Following discovery of the failure, the Unit 2 HPCS injection isolation valve 2E22-F004 was overhauled using a new stem, and the upper wedge threads were repaired. In addition, a stem lube and rotation check was performed satisfactorily. A review of the diagnostic testing results and pin analysis criteria was performed on all the affected valves in this population. The remaining valves in the population were within the latest vendor's recommendation for rotation criteria.
HPCS injection isolation valve 1E22-F004 internals were replaced during a maintenance outage performed in late June 2017.
PREVIOUS OCCURRENCES
A review of station Licensee Event Reports for the past three years, related to stem-disc separation issues, identified the following similar instances:
neutron flux, followed by a Group I containment isolation. Following the Group I isolation, the control room operators noted that the position indication for valve 2B21-F022C, the inboard 2C Main Steam Isolation Valve (MSIV), showed dual indication rather than full closed. Troubleshooting of the 2C MSIV determined that the valve stem disc had separated from the stem, which allowed the main disk to drop into the main steam flow path. The resulting reactor pressure transient added positive reactivity, which caused the high neutron flux scram. Increased steam flow in the other three main steam lines resulted in a nearly simultaneous high main steam line flow Group I containment isolation. The cause of the stem-disc separation on the 2C MSIV was fretting wear attributable to marginal design. The root cause of the event was a legacy decision made in 2008 deferring installation of a comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Ccmmission, Washington, DC 20555-0001, or by e-mail to ilniccollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, used to impose an information collection dces not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
3. LER NUMBER
2017 - 01 003 manufacturer upgrade that would have prevented the failure. Corrective actions include installing the upgrade on all MSIVs on botli units, and reviewing previous deferral decisions made using the same decision-making process.
result of observing a generator run-back due to a generator stator winding cooling (GC) system malfunction. Initial troubleshooting identified the most likely cause was plugging in the 'A' GC heat exchanger, based on inspection of the GC system flow-path components. The GC system was realigned to the 'B' heat exchanger until inspections could be performed in the upcoming refueling outage, and the unit was re-started on January 24, 2017. Further inspections of the GC components were performed while the unit was shut down for a planned refueling outage. These inspections determined the cause of the GC system failure was stem-disc separation in the 'A' GC heat exchanger inlet valve. The valve was repaired during the refueling outage. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the Reactor Protection System (RPS). There were no safety consequences associated with the event since there was no loss of safety function, and the RPS functioned as designed.
Diesel Generator Cooling Water (DGCW) system, the cooling water strainer backwash valve was unable to open due to stem-disc separation. The valve was replaced, and the HPCS system was returned to operable on February 2, 2017. This condition could have prevented the HPCS system, a single train safety system, from performing its design function. There was minimal safety consequences associated with the event since the other emergency safety systems remained operable, and the Division 3 DGCW system remained functional as it retained the ability to provide the required flow through the system. The apparent cause of the stem-disc separation was erosion due to the carbon-steel valve internals in a raw water system environment.
COMPONENT FAILURE DATA
Manufacturer: Anchor Darling (A391) Device: Gate Valve, 12-inch Component ID: Model C900