On February 23, 2017, radiography results for the Browns Ferry Nuclear Plant ( BFN), Unit 2, Reactor Core Isolation Cooling ( RCIC) system stop check valve, 2-HCV-071-0014 (71-14 valve), showed the valve to be in the fully open position.
In this position, the valve was incapable of performing its design function as a Primary Containment Isolation Valve (PCIV).
In order to isolate the affected primary containment penetration flowpath, the RCIC Steam Line Outboard Isolation Valve was closed and deactivated to secure flow through the adjacent PCIV check valve in the exhaust line in accordance with plant Technical Specifications (TS) Limiting Conditions for Operation (LCO) for Primary Containment Isolation.
It is assumed in the Past Operability Evaluation (POE) that the 71-14 valve had been inoperable from April 2015 to February 2017, in violation of TS LCO 3.6.1.3. During this time period, another check valve (upstream of the 71-14 valve) remained capable of isolating the affected penetration. The 71-14 valve was repaired during the BFN, Unit 2, Cycle 19 Refueling Outage which began on February 25, 2017.
The cause of this condition was determined to be inadequate minimum allowable internal clearances for the 71-14 valve.
The 71-14 valve internals were machined to minimum allowable material as an interim action to mitigate risks. The corrective action to prevent recurrence is to replace the 71-14 valve with a new valve design that is not susceptible to the same internal binding potential. |
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Category:Letter
MONTHYEARML24032A4762024-02-0101 February 2024 Final Report of a Part 21 Evaluation Associated with Starter Contactors for the BFN Unit 1 High Pressure Coolant Injection Suppression Pool Inboard Suction Valve ML24023A2802024-01-23023 January 2024 Final Report of a Deviation or Failure to Comply Associated with a Relay in the Reactor Core Isolation Cooling Condensate Pump CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions ML24016A3042024-01-16016 January 2024 Final Report of a Part 21 Evaluation Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump CNL-23-071, Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, Inservice Inspection and Augmented Program Interval Revised Request for Alternative 0-ISI-472024-01-11011 January 2024 Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, Inservice Inspection and Augmented Program Interval Revised Request for Alternative 0-ISI-47 ML24022A1732024-01-0303 January 2024 Receipt and Availability of the Subsequent License Renewal Application ML23319A1992024-01-0303 January 2024 Issuance of Amendment Nos. 333, 356, and 316 Regarding the Technical Specification Surveillance Requirements 3.4.3.2 and 3.5.1.11 Regarding Safety Relief Valves ML23355A2062023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23348A3942023-12-14014 December 2023 Interim Part 21 Report of a Potential Deviation or Failure to Comply Associated with Starter Contactors for the High Pressure Coolant Injection Suppression Pool Inboard Suction Valve IR 05000259/20230102023-12-11011 December 2023 Commercial Grade Dedication Inspection Report 05000259/2023010 and 05000260/2023010 and 05000296/2023010 ML23335A0722023-12-0101 December 2023 Interim Report of a Deviation or Failure to Comply Associated with a Relay in the Unit 2 Reactor Core Isolation Cooling Condensate Pump ML23334A2492023-11-30030 November 2023 Site Emergency Plan Implementing Procedure Revision CNL-23-070, Submittal of Fifth 10-Year Interval Inservice Testing Program Plan2023-11-29029 November 2023 Submittal of Fifth 10-Year Interval Inservice Testing Program Plan ML23331A2532023-11-27027 November 2023 Summary Report for 10 CFR 50.9 Evaluations, Technical Specifications Bases Changes, Technical Requirement Manual Changes, and NRC Commitment Revisions CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23325A1102023-11-21021 November 2023 Anchor Darling Double Disc Gate Valve Commitment Revision ML23320A2542023-11-16016 November 2023 Interim Part 21 Report of a Deviation or Failure to Comply Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump IR 05000259/20230032023-11-13013 November 2023 Integrated Inspection Report 05000259/2023003, 05000260/2023003 and 05000296/2023003 IR 05000259/20230402023-11-0202 November 2023 Supplemental Inspection Supplemental Report 05000259 2023040 and Follow-Up Assessment Letter ML23292A2532023-10-18018 October 2023 BFN 2024-301, Corporate Notification Letter (210-day Ltr) ML23282A0022023-10-0606 October 2023 Interim Part 21 Report of a Deviation or Failure to Comply Associated with Check Valve 0-CKV-023-0565 for D1 Residual Heat Removal Service Water Pump ML23278A0122023-10-0505 October 2023 Updated Final Safety Analysis Report, Amendment 30 ML23271A1702023-09-28028 September 2023 Site Emergency Plan Implementing Procedure Revision ML23270A0702023-09-26026 September 2023 SLRA Pre-Application Meeting Summary 09-13-2023 ML23257A1232023-09-22022 September 2023 Administrative Changes to Technical Specification Pages Issued for License Amendment Nos. 332, 355, and 315 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision ML23263B1042023-09-20020 September 2023 Special Report 260/2023-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation ML23205A2132023-09-0808 September 2023 Issuance of Amendment Nos. 332, 355, and 315 Regarding the Revision of Technical Specifications to Adopt TSTF-566-A and TSTF-580-A, Rev. 1 CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000259/20230052023-08-29029 August 2023 Updated Inspection Plan for Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Report 05000259/2023005, 05000260/2023005 and 05000296/2023005 ML23233A0432023-08-18018 August 2023 Enforcement Action EA-22-122 Inspection Readiness Notification ML23219A1542023-08-17017 August 2023 Request to Use Later Edition of ASME Code for Operation and Maintenance and Alternative Requests BFN-IST-01 Through 05 for the Fifth 10-Year Interval Inservice Testing Program ML23228A1642023-08-16016 August 2023 Site Emergency Plan Implementing Procedure Revision ML23228A0202023-08-15015 August 2023 (BFN) Unit 1 - Special Report 259/2023-001 for Inoperable Post Accident Monitoring (PAM) Instrumentation IR 05000259/20230022023-08-10010 August 2023 Integrated Inspection Report 05000259/2023002, 05000260/2023002, 05000296/2023002 and 07200052/2023001 ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information CNL-23-055, Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills2023-07-25025 July 2023 Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills ML23171A8862023-07-24024 July 2023 Issuance of Amend. Nos. 331, 354, and 314; 365 and 359 Regarding Adoption of TSTF-541-A, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position ML23201A2182023-07-20020 July 2023 Registration of Use of Cask to Store Spent Fuel (MPC-298 and -299) ML23159A2552023-07-20020 July 2023 Proposed Alternative to the Requirements of the ASME Code Regarding Volumetric Inspection of Standby Liquid Control Nozzles ML23199A3072023-07-18018 July 2023 Site Emergency Plan Implementing Procedure Revision CNL-23-053, Tennessee Valley Authority - Radiological Emergency Plan Revisions2023-07-18018 July 2023 Tennessee Valley Authority - Radiological Emergency Plan Revisions IR 05000259/20233012023-07-18018 July 2023 NRC Operator License Examination Report Nos. 05000259/2023301, 05000260/2023301, and 05000296/2023301 2024-02-01
[Table view] Category:Licensee Event Report (LER)
MONTHYEARML20160A0232020-06-0404 June 2020 SR 2020-001-00 for Browns Ferry Nuclear Plant (Bfn),Inoperable Oscillating Power Range Monitor (OPRM) Instrumentation 05000296/LER-2017-0022017-12-29029 December 2017 4kV Shutdown Board Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses, LER 17-002-00 for Browns Ferry Nuclear Plant, Unit 3 Regarding 4kV Shutdown Board Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses 05000296/LER-2017-0012017-10-31031 October 2017 Inoperable Residual Heat Removal Pump Results in Condition Prohibited by Technical Specifications, LER 17-001-00 for Browns Ferry Nuclear Plant, Unit 3, Regarding Inoperable Residual Heat Removal Pump Results in Condition Prohibited by Technical Specifications 05000260/LER-2017-0042017-07-0707 July 2017 Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints, LER 17-004-00 for Browns Ferry, Unit 2, Regarding Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints 05000260/LER-2017-0032017-05-30030 May 2017 Manual Reactor Scram Initiated During Startup Due to Multiple Rods Inserting, LER 17-003-00 for Browns Ferry Nuclear Plant, Unit 2 Regarding Manual Reactor Scram Initiated During Startup Due to Multiple Rods Inserting 05000259/LER-2017-0022017-04-27027 April 2017 Unauthorized Firearm Introduced into the Protected Area, LER 17-002-00 for Browns Ferry, Unit 1, Regarding Unauthorized Firearm Introduced into the Protected Area 05000260/LER-2017-0022017-04-24024 April 2017 Inoperable Primary Containment Isolation Valve Resulting in Condition Prohibited by Technical Specifications, LER 17-002-00 for Browns Ferry, Unit 2, Regarding Inoperable Primary Containment Isolation Valve Resulting in Condition Prohibited by Technical Specifications 05000260/LER-2017-0012017-04-14014 April 2017 High Pressure Coolant Injection Safety System Functional Failure Due to a Blown Fuse, LER 17-001-00 for Browns Ferry, Unit 2, Regarding High Pressure Coolant Injection Safety System Functional Failure Due to a Blown Fuse 05000259/LER-2016-0022016-09-19019 September 2016 High Pressure Coolant Injection Safety System Functional Failure due to Inoperability of Primary Containment Isolation Valve, LER 16-002-00 for Browns Ferry, Unit 1, Regarding High Pressure Coolant Injection Safety System Functional Failure Due to Inoperability of Primary Containment Isolation Valve 05000260/LER-2016-0022016-09-13013 September 2016 High Pressure Coolant Injection System Failure Due To Stuck Contactor, LER 16-002-00 for Browns Ferry Nuclear Plant, Unit 2, Regarding High Pressure Coolant Injection System Failure Due To Stuck Contactor 05000260/LER-2016-0012016-08-16016 August 2016 High Pressure Coolant Injection Safety System Functional Failure due to a Blown Fuse and a Failed Relay, LER 16-001-00 for Browns Ferry, Unit 2, Regarding High Pressure Coolant Injection Safety System Functional Failure Due to a Blown Fuse and a Failed Relay 05000296/LER-2016-0062016-08-0505 August 2016 1 OF 8, LER 16-006-00 for Browns Ferry Nuclear Plant, Unit 3, Regarding High Pressure Coolant Injection System Found to be Inoperable During Testing 05000259/LER-2016-0012016-06-21021 June 2016 Failure of 4kV Shutdown Board Normal Feeder Breaker Results in Actuations of Emergency Diesel Generators and Containment Isolation Valves, LER 16-001-00 for Browns Ferry, Unit 1, Regarding Failure of 4kV Shutdown Board Normal Feeder Breaker Results in Actuations of Emergency Diesel Generators and Containment Isolation Valves 05000296/LER-2016-0052016-06-17017 June 2016 Automatic Depressurization System Valve Inoperability Exceeded Technical Specification Limits, LER 16-005-00 for Browns Ferry, Unit 3, Regarding Automatic Depressurization System Valve Inoperability Exceeded Technical Specification Limits 05000296/LER-2016-0042016-06-0606 June 2016 Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints, LER 16-004-00 for Browns Ferry, Unit 3, Regarding Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints 05000296/LER-2016-0032016-04-25025 April 2016 Main Steam Isolation Valve Leaking in Excess of Technical Specification Requirements, LER 16-003-00 for Browns Ferry Nuclear Plant Unit 3 Regarding Main Steam Isolation Valve Leaking in Excess of Technical Specification Requirements 05000296/LER-2016-0022016-04-22022 April 2016 Improperly Installed Switch Results in Condition Prohibited by Technical Specifications, LER 16-002-00 for Browns Ferry Nuclear Plant, Unit 3, Regarding Improperly Installed Switch Results in Condition Prohibited by Technical Specifications 05000296/LER-2016-0012016-03-21021 March 2016 Inoperable Residual Heat Removal Pump Results in Condition Prohibited by Technical Specifications and Safety System Functional Failure, LER 16-001-00 for Browns Ferry, Unit 3, Regarding Inoperable Residual Heat Removal Pump Results in Condition Prohibited by Technical Specifications and Safety System Functional Failure 05000260/LER-2015-0022016-03-17017 March 2016 High Pressure Coolant Injection System Inoperable due to Manual Isolation of Steam Leak I, LER 15-002-01 for Browns Ferry, Unit 2, Regarding High Pressure Coolant Injection System Inoperable Due to Manual Isolation of Steam Leak ML1108400352011-03-22022 March 2011 Letter Re Licensee Event Report Which Occurred on December 22, 2010, Concerning Low Pressure Coolant Injection Operability, TVA Expects to Submit a Revised LER by April 15, 2011 ML1015505752010-04-0707 April 2010 Event Notification for Browns Ferry on Spill of Water Containing Tritium ML1015505632008-01-10010 January 2008 Event Notification for Browns Ferry on Offsite Notification - Spill of Water Containing Tritium ML18283B3261978-09-29029 September 1978 LER 1978-205-01 for Browns Ferry, Unit 3 Four Main Steam Isolation Valves Which Exceeded the Leakage Limits of Technical Specification 4.7.A.2.i While Performing Local Leak Rate Testing During Refueling ML18283B3391978-07-25025 July 1978 Licensee Event Report Concerning Excessive Drywell Floor Drain Leak Rate Observed During Normal Operation ML18283B3411978-07-18018 July 1978 Licensee Event Report Concerning an Abnormal Indication on a 4-kV Standby Power Circuit Breaker During Normal Operation ML18283B3401978-07-18018 July 1978 Licensee Event Report Concerning an Outboard Main Steam Isolation Valve, Which Closed Faster than Allowed by Technical Specifications ML18283B3421978-05-31031 May 1978 Licensee Event Report Concerning MSIV 1-38 Which Closed in 1 Second Exceeding Limiting Condition of Operation ML18283A9901978-05-30030 May 1978 LER 1978-010-00 for Browns Ferry Nuclear Plant, Unit 2, Relief Valve on Standby Liquid Control Pump B Opened at 900 Psig (Which Is Lower than Designed Setting of 1425 +/- 75 Psig as Designated by Tech Spec 4.4.A.2.A) During Surveillance Tes ML18283A9911978-05-0909 May 1978 LER 1978-008-00 for Browns Ferry Nuclear Plant, Unit 2, Reactor Building Ventilation Radiation Monitoring Channel Failed During Refueling Outage ML18283A9941978-05-0505 May 1978 LER 1978-009-00 for Browns Ferry Nuclear Plant, Unit 2, Local Leak Rate Tests of All Containment Isolation Valves Where Leak Rate Exceeded Allowable Leak Rate of 60 Percent of La Per 24 Hours or 707.1 Scfh ML18283A9921978-05-0505 May 1978 LER 1978-006-00 for Browns Ferry Nuclear Plant, Unit 2, Check Valve 2-73-603 in High-Pressure Coolant Injection System Was Found in Open Position During Maintenance Inspection After Failing Local Leak Rate Test ML18283B4001978-05-0101 May 1978 LER 1978-010-00 for Browns Ferry Nuclear Plant, Unit 3, Both RBM Channels Which Became Continuously Bypassed During Power Ascension ML18283B4011978-04-28028 April 1978 LER 1978-009-00 for Browns Ferry Nuclear Plant, Unit 3, Smoke Alarm Which Would Not Clear & Was Received for Preaction Sprinkler Zone in Reactor Building During Normal Operation ML18283B4021978-04-28028 April 1978 LER 1978-008-00 for Browns Ferry Nuclear Plant, Unit 3, Relief Valve 3-1-31 Which Failed to Reseat Until Reactor Pressure Reached 280 Psig During Reactor Scram ML18283B4041978-04-24024 April 1978 LER 1978-001-00 for Browns Ferry Nuclear Plant, Unit 3, Torus Oxygen Sensor O2M-76-42, Found to Be Erratic & Did Not Meet Requirements of Tech Spec 4.7.II During Normal Operation, Which Is Superseding Previous Letter of 2/8/1978 ML18283B4031978-04-24024 April 1978 LER 1978-006-00 for Browns Ferry Nuclear Plant, Unit 3, Electrical Connector Carrying Thermocouple Circuits Monitoring Primary Containment Atmospheric Temperature Not Included as Part of Modification Which Qualified Connector Assemblies for ML18283B4051978-04-0404 April 1978 LER 1978-005-00 for Browns Ferry Nuclear Plant, Unit 3, Six CRD Accumulator Level Switches Which Would Not Alarm with Level Increases During Plant Operation While Performing Electrical Maintenance Instruction 50 ML18283B4061978-03-30030 March 1978 LER 1977-012-00 for Browns Ferry Nuclear Plant, Unit 3, Temperature Transients Which Were Experienced with Six Charcoal Adsorber Beds in Offgas System During Normal Operation, Which Is Supplementing Previous Letter of 7/29/1977 ML18283A9951978-03-29029 March 1978 LER 1978-005-00 for Browns Ferry Nuclear Plant, Unit 2, Unidentified Coolant Leakage in Drywall Was Found to Be 9.5 Gpm & Exceeded 5 Gpm Limit of Technical Specification 3.6.C.1. During Normal Operation ML18283B4091978-03-28028 March 1978 LER 1978-004-00 for Browns Ferry Nuclear Plant, Unit 3, Three of Five Test Specimens Failed During Simulated LOCA Conditions & During Qualification Testing of Bendix Electrical Connectors Identical to Those Used in Primary Containment ML18283B4101978-03-22022 March 1978 LER 1977-005-00 for Browns Ferry Nuclear Plant, Unit 3, RPS MG Set a Which Continued Running & MG Set B Output Breaker Which Did Not Trip During Startup Test STI-31, Which Is Supplementing Previous Letter of 3/24/1977 ML18283B4111978-03-10010 March 1978 LER 1978-003-00 for Browns Ferry Nuclear Plant, Unit 3, Valve FCV 3-74-52 Was Found Inoperable During Performance of Surveillance Instruction 4.5.B.1.C ML18283A9961978-02-28028 February 1978 LER 1978-004-00 for Browns Ferry Nuclear Plant, Unit 2, Relief Valve 2-1-5 Opened and Failed to Reseat During Steady State Operation ML18283B4141978-02-28028 February 1978 LER 1978-002-00 for Browns Ferry Nuclear Plant, Unit 3, Bendix Connectors of Type Used Inside Primary Containment Have Failed a Post-Aging Environmental Test at Wyle Laboratory Testing Facility ML18283A9971978-02-15015 February 1978 LER 1978-003-00 for Browns Ferry Nuclear Plant, Unit 2, Relief Valve 2-1-41 Opened and Failed to Reseat During Steady State Operation ML18283B0001978-02-13013 February 1978 LER 1978-002-00 for Browns Ferry Nuclear Plant, Unit 2, Reactor Pressure Switch PS-68-95 Not Functioning as Required by Tech Spec Table 4.2.B During Normal Operation While Performing Surveillance Instruction ST 4.2.B-7 ML18283A9991978-02-0606 February 1978 LER 1978-001-00 for Browns Ferry Nuclear Plant, Unit 2, Surveillance Samples Were Taken from Charcoal in Unit 2 Primary Containment Purge System Following Maintenance Outage ML18283B4161978-02-0606 February 1978 LER 1978-001-00 for Browns Ferry Nuclear Plant, Unit 3, Torus Oxygen Sensor O2M-76-42 Found to Be Erratic & Did Not Meet Requirements of Technical Specification 4.7.H During Normal Operation ML18283B4071977-10-0505 October 1977 LER 1977-012-00 for Browns Ferry Nuclear Plant, Unit 3, Temperature Transients Were Experienced with Six Charcoal Adsorber Beds in Offgas System ML18283B4171977-09-26026 September 1977 LER 1977-017-00 for Browns Ferry Nuclear Plant, Unit 3, Primary Containment Isolation Valve 3-FCV-77-2A on Drywell Floor Drain Sump Pump Discharge Line Would Not Operate as Required by Tech Spec 3.7.D.L During Routine Operability Checks 2020-06-04
[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.
I. Plant Operating Conditions Before the Event
At the time of discovery, Browns Ferry Nuclear Plant (BFN), Unit 2, was in Mode 1 at approximately 74 percent rated thermal power.
II. Description of Event
A. Event Summary:
On February 23, 2017, at approximately 2239 Central Standard Time (CST), radiography results for BFN, Unit 2, Reactor Core Isolation Cooling (RCIC) [BN] system stop check valve [ISV], 2-HCV-071-0014 (71-14 valve), was discovered to be in the fully open position, thereby not meeting the acceptance criteria of Surveillance Instruction 2-Sl-3.2.3, Testing ASME OM Code and Augmented IST Check Valves (Internal Inspection). This valve is the RCIC turbine exhaust stop check valve and is one of two Primary Containment Isolation Valves (PCIVs) connecting the RCIC exhaust directly to primary containment. The acceptance criteria is for the valve to be fully closed to satisfy the containment isolation function. A Past Operability Evaluation (POE) determined that the 71-14 valve was inoperable from April 3, 2015, at 2356 Central Daylight Time (CDT), until Unit 2 was placed in Mode 4 on February 25, 2017, at 0506 CST.
BFN, Unit 2, Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.6.1.3 requires each PCIV, except reactor building-to-suppression chamber vacuum breakers, to be Operable in Modes 1, 2, and 3, and when associated instrumentation is required to be Operable per LCO 3.3.6.1, “Primary Containment Isolation Instrumentation”. In response to this event, on February 23, 2017, at 2250 CST, Operations personnel entered TS 3.6.1.3 Required Action A.1 which required the RCIC system to be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to prevent flow through the RCIC Turbine Exhaust check valve, 2-CKV-071-0580 (71-580 valve). Based on the results from the POE, the 71-14 valve was inoperable longer than allowed by the TS. On February 24, 2017, at 0018 CST, Operations personnel isolated the RCIC system in accordance with plant TS by closing the RCIC Steam Line Outboard Isolation Valve 2-FCV-071-0003 (71-3 valve).
On February 25, 2017, the scheduled Unit 2 refueling outage commenced.
On March 24, 2017 at 1641 CDT, BFN completed repairs to the 71-14 valve and restored the RCIC penetration flowpath.
B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event The 71-14 valve was unable to perform its function as a PCIV because it was stuck in the open position.
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.
C. Dates and approximate times of occurrences
Dates & Approximate Times Occurrence April 3, 2015, at 2356 CDT It is assumed in the POE that the 71-14 valve bound open immediately after the start of the RCIC Turbine during performance of 2-SR-3.5.3.4, RCIC System Rated Flow at Low RPV Pressure.
February 23, 2017, at 2239 CST Radiography results for the 71-14 valve showed the valve to be in the fully open position.
February 23, 2017, at 2250 CST Operations declared 71-14 valve inoperable and entered TS Action 3.6.1.3.A to secure flow through check valve 71-580.
February 24, 2017, at 0018 CST Valve 71-3 was closed to secure flow through check valve 71-580 which also removed the RCIC system from service.
February 25, 2017, at 0001 CST Unit 2 inserted a manual reactor scram and entered Mode 3 for the previously scheduled Unit 2 refueling outage.
February 25, 2017, at 0506 CST Unit 2 entered Mode 4 where operability of the 71-14 valve was no longer required.
March 24, 2017, at 1641 CDT Operations personnel restored the RCIC penetration flowpath following completion of repairs to the 71-14 valve.
D. Manufacturer and model number of each component that failed during the event The failed component was the 71-14 valve. The component was model 5312WE, manufactured by the Walworth Company.
E. Other systems or secondary functions affected
The 71-14 valve is one of two PCIVs connecting the RCIC exhaust directly to primary containment.
Containment was isolated by closing the 71-580 valve resulting in the RCIC system being unable to perform its design function.
III. Cause of the event
A. Cause of each component or system failure or personnel error The root cause of this condition was inadequate minimum allowable clearances for the 71-14 valve.
There were two direct causes identified for the valve's failure:
F. Method of discovery of each component or system failure or procedural error On February 23, 2017, radiography results for the BFN, Unit 2, 71-14 valve showed the valve to be in the fully open position.
G. The failure mode, mechanism, and effect of each failed component The disc of the 71-14 valve was stuck in the fully open position due to internal binding. Three factors aligned to cause the internal binding of the valve: limited internal clearance for moving parts, corrosion debris from the RCIC piping was deposited into the valve, and the valve bonnet was not centered on the body. This configuration resulted in the 71-14 valve being unable to perform its safety function as a PCIV.
H. Operator actions
Operations personnel declared the 71-14 valve inoperable. The Primary Containment flowpath was isolated by securing the 71-580 valve. The 71-3 valve was closed, resulting in the RCIC system becoming inoperable.
I. Automatically and manually initiated safety system responses
2. The 71-14 valve was mechanically bound, preventing the valve disc from returning to the closed position.
B. Cause(s) and circumstances for each human performance related root cause There were no human performance root causes.
1. The Corrective Action to Prevent Recurrence developed for a similar 2013 valve failure, which was to submit a Preventive Maintenance Change Request to open, inspect, clean, and replace valve disc, dashpot and stem as necessary for Units 1, 2, and 3 71-14 valves, did not prevent recurrence of future similar events.
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3. LER NUMBER
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.
05000-260
IV. Analysis of the event
The Tennessee Valley Authority (TVA) is submitting this report in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(i)(B), as any operation or condition which was prohibited by the plant's TS.
The 71-14 valve is one of two PCIVs connecting the RCIC exhaust directly to primary containment through penetration X-212. This 71-14 valve has a safety-related function in the closed position and a quality-related, important to safety function, in the open position. According to TS Bases 3.6.1.3 for PCIVs, PCIVs form a part of the primary containment boundary. The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a Design Basis Accident. PCIV operability supports leak tightness of primary containment.
The 71-14 valve is normally closed to provide its safety-related function as containment isolation. The valve opens from its normally closed position to allow a flow path from the RCIC Turbine Exhaust to the Pressure Suppression Chamber (PSC). The 71-14 valve closes when the RCIC system is shut down due to the weight of the valve disc. The exhaust header is designed to contain a check valve and a stop check valve to prevent backflow of Suppression Pool water into the RCIC exhaust line.
TS LCO 3.6.1.3 for PCIVs requires each PCIV, except reactor building-to-suppression chamber vacuum breakers, to be Operable in Modes 1, 2, and 3. However, on February 23, 2017, the 71-14 valve was found failed open during surveillance testing. At the time of discovery, Unit 2 was in Mode 1. As a result of the 71-14 valve being declared inoperable, Operations personnel entered TS 3.6.1.3 required actions to isolate the RCIC system preventing flow through the 71-580 valve which is the upstream RCIC turbine exhaust line PCIV. The 71-14 valve was determined to have been inoperable longer than allowed by TS resulting in a condition prohibited by the plant's TS.
V. Assessment of Safety Consequences
The ability of the Primary Containment penetration X-212 to perform its design function was not adversely affected by the 71-14 valve being bound in the open position due to the 71-580 valve remaining capable of isolating the penetration.
Additionally, the ability of the 71-14 valve to fulfill its function of providing a flow path for RCIC Turbine Exhaust Steam was not adversely affected by it being bound in the open position. Since the 71-14 valve must be in the open position to operate RCIC, the RCIC System injection function was not affected when this valve was bound open; therefore, the RCIC System was capable of injecting water into the reactor vessel and would have met its mission time.
Therefore, the RCIC system was still capable of performing its required safety functions to protect the 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.
health and safety of the public.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event With the 71-14 valve being bound in the open position, this valve could not perform its design function as a PCIV. However, the 71-580 valve remained capable of isolating the penetration.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident This event did not occur when Unit 2 reactor was shutdown.
C. For failure that rendered a train of a safety system inoperable, estimate of the elapsed time from discovery of the failure until the train was returned to service It is assumed in the POE that the 71-14 valve was inoperable from April 3, 2015, at 2356 CDT, until Unit 2 was placed in Mode 4 on February 25, 2017, at 0506 CST.
VI. Corrective Actions
Corrective Actions are being managed by TVA's corrective action program under Condition Report (CR) 1265552.
A. Immediate Corrective Actions
The 71-14 valve was repaired.
B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future The root cause of this event will be addressed by replacing the Unit 2 71-14 valve with a new valve design that is not susceptible to the same internal binding potential as the existing valve.
VII. Previous Similar Events at the Same Site Previous similar events at BFN include failure of the Unit 2 71-14 valve. On March 2, 2013, radiography results for the Unit 2 71-14 valve displayed the valve bound in the fully open position. This event was reportable, and LER 50-260/2013-001-00, Inoperable Stop Check Valve Results in Condition Prohibited by Technical Specifications and Reactor Shutdown Required by Technical Specifications, was submitted to the Nuclear Regulatory Commission on May 1, 2013.
A search was performed of CRs relating to the 71-14 valve. CRs related to the condition reported in the
VIII. Additional Information
There is no additional information.
IX. Commitments There are no new commitments.
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.
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