06-17-2016 | On April 18, 2016, during a scheduled surveillance, the power to the Main Steam Line ( MSL) B Relief Valve failed to transfer to its alternate feeder breaker when the normal feeder breaker was opened. The Automatic Depressurization System ( ADS) function of the MSL B Relief Valve was declared inoperable. It was determined that the ADS valve was inoperable from March 26, 2016 to April 19, 2016. The valve's ability to open under normal power was not affected. Five of the six ADS valves remained operable. Only four ADS valves are required to meet the ADS function in the Loss of Coolant Analysis described in the Final Safety Accident Report.
The unavailability of the ADS alternate power source was directly caused by a bus stab on the back of the Molded Case Circuit (MCC) breaker not fully engaging with the bus. This was apparently caused by improper performance of previous post-maintenance testing. The stab was adjusted, the MCC breaker was returned to service, and the MSL B Relief Valve's ADS function was declared operable upon verification of its alternate power supply.
Corrective actions were to determine which load-feeding MCC breakers have a normal and alternate power source, and revise their preventative maintenance procedures to verify that post-maintenance testing includes power source isolation prior to closing the breaker under load. Breaker bus stabs will be replaced. |
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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 FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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 3, was in Mode 1 at 100 percent power.
II. Description of Events
A. Event:
On April 18, 2016, during a scheduled surveillance of the Automatic Depressurization System (ADS) [SB] Logic System, the power to the Main Steam Line (MSL)[SB] B Relief Valve [RV] (3-PCV-001-0022) failed to transfer to the alternate feeder breaker when the normal feeder breaker was opened. Operations subsequently declared the ADS function of the MSL B Relief Valve inoperable, and entered Technical Specification (TS) Limiting Condition of Operability (LCO) 3.5.1 Condition E, a 14 day shutdown LCO action statement. Troubleshooting determined that the bus stab on the back of the alternate feeder Molded Case Circuit (MCC) breaker [BKR] (3-BKR-001-0022A) was not properly engaged with the bus. The stab was adjusted, the MCC breaker was returned to service, and MSL B Relief Valve was verified to have proper power. The valve's ADS function was declared operable on April 19, 2016 at 1400 Central Daylight Time (CDT).
Since the failure was discovered during surveillance testing, a past operability evaluation (POE) was performed to determine the duration of inoperability. The POE determined that the breaker was working properly until it was racked out for preventative maintenance (PM) performed during the previous Unit 3 refueling outage in March 2016. This failure went undetected due to improper post-maintenance testing (PMT). Prior to the outage, the associated breaker passed all required surveillances. The POE concluded that the time of inoperability began on March 26, 2016 at 2300 CDT and ended on April 19, 2016 at 1400 CDT.
Since the BFN, Unit 3 ADS system only requires four of the six ADS valves to remain operable, the ADS system was operable throughout this entire event. MSL B Relief Valve actuation through the ADS or Main Steam Relief Valve (MSRV) Automatic Actuation Logics were not affected under normal power. The valve's mechanical setpoints were unaffected by this event, and they remained capable of lifting.
B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event:
There were no structures, systems, or components (SSCs) whose inoperability contributed to this event.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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.
3. LER NUMBER
2016 - 00
C. Dates and approximate times of occurrences:
Dates and Approximate Times Occurrence April 22, 2014 Last successful performance of 3-SR-3.3.5.1.6(ADS A), ADS Logic System Functional Test.
March 9, 2016 Performed breaker inspection and PM on 250V DC Motor- Operated Valve Board 3A.
March 26, 2016 at 2300 CDT Plant mode and conditions require ADS valve operability.
April 18, 2016 April 19, 2016 at 1400 CDT During scheduled performance of 3-SR-3.3.5.1.6(ADS A), Power to the MSL B Relief Valve failed to transfer to its alternate feeder breaker when the normal feeder breaker was opened. The ADS valve was declared inoperable.
Troubleshooting found a stab on the back of the MCC breaker which required adjustment. Following adjustment, the ADS valve was returned to operable status.
D. Manufacturer and model number (or other identification) of each component that failed during the event:
The inoperable valve was a Target Rock Corporation two-stage pressure control valve, model number 7567F. The breaker was manufactured by General Electric, model number THEF136015.
E. Other systems or secondary functions affected:
No other systems or secondary functions were affected by this event.
F. Method of discovery of each component or system failure or procedural error:
Valve inoperability was discovered during the scheduled performance of 3-SR-3.3.5.1.6(ADS A), ADS Logic System Functional Test, when Operations opened the normal feeder breaker for the MSL B Relief Valve on 250V DC Motor-Operated Valve Board 3A, and its associated green light did not remain illuminated. Breaker inoperability was discovered during troubleshooting for the inoperable ADS valve.
G. The failure mode, mechanism, and effect of each failed component, if known:
The MSL B Relief Valve failed when its associated MCC breaker, which supplied alternate power from 250V DC Motor-Operated Valve Board 3A, became unavailable. The MCC breaker failed because the bus stabs were not properly engaged with the bus.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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.
3. LER NUMBER
2016 - 00
H. Operator actions:
There were no operator actions associated with this event.
I. Automatically and manually initiated safety system responses:
There were no automatic or manual safety system responses associated with this event.
Ill. Cause of the event A. The cause of each component or system failure or personnel error, if known:
The direct cause for the ADS valve failure was the unavailability of its alternate power source due to a bus stab on the back of the MCC breaker not fully engaging with the bus. The most apparent cause for the valve failure was improper implementation of previous PMT.
B. The cause(s) and circumstances for each human performance related root cause:
Previous PMT was not properly performed, and failed to check if load operation was transferred upon breaker closure.
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 50.73(a)(2)(i)(B), as any operation or condition which was prohibited by the plant's TS. It was determined that the ADS valve was inoperable from March 26, 2016 to April 19, 2016, due the failure of the valve's associated MCC breaker to supply alternate power because of a misaligned bus stab.
BFN, Unit 3, TS LCO 3.5.1 requires six Operable ADS Safety/Relief Valves (S/RVs) during Modes 1, and in Modes 2 and 3 when reactor steam dome pressure exceeds 150 psig. If one or more required ADS valve becomes inoperable, Required Action E.1 requires that valve to be restored to operability within 14 days. Based on this evaluation, BFN, Unit 3, operated with an inoperable ADS valve for longer than allowed by TS.
The condition was assumed to have begun on March 9, 2016, during routine breaker inspection, as part of refueling outage maintenance activities. ADS valve operability was not required until March 26, 2016, when BFN, Unit 3 entered Mode 2 and reactor dome pressure exceeded 150 psig.
Therefore, the ADS valve was considered inoperable from March 26, 2016 to April 19, 2016, when the breaker stabs were readjusted. The duration of ADS system inoperability was longer than allowed by plant TS LCO 3.5.1.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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.
3. LER NUMBER
2016 - 00
V. Assessment of Safety Consequences
System availability was not impacted by this event. While one ADS valve was inoperable, the remaining five ADS valves remained operable for the duration of this event. Only four ADS valves are required to meet the ADS function in the Loss of Coolant Accident analysis described in the Final Safety Accident Report. This condition did not affect the Appendix R Function. However, due to the failure of its alternate power source, the MSL B Relief Valve was not able to perform its specified safety function under all conditions.
The MSL B Relief Valve remained operable for its required "Overpressure safety and relief protection for the nuclear system" function. The lack of reliable alternate power, from the MCC breaker had no effect on the mechanical setpoint of the valve, and it would have opened if reactor pressure exceeded 1145 psig.
The MSL B Relief Valve was inoperable for its "Automatic nuclear system depressurization" function.
All ADS valves are equipped with a logic circuit to automatically open the valve on concurrent high drywell pressure and low reactor water level or sustained reactor low water level when one of the Residual Heat Removal pumps is available in the Low Pressure Coolant Injection mode or two of the appropriate core spray pumps are available. Additionally, four of the six ADS valves have alternate power sources for manual actuation. The MSL B Relief Valve normal power source was aligned and operating as designed. However, during testing it was found that the alternate power source was not working as designed. The fuel vendor analyses of record assume between four and six available ADS valves, depending upon the applicable assumed single failure. A single failure of 250V DC Motor- Operated Valve Board 3A, which feeds the MSL B Relief Valve normal power supply would still have resulted in all six S/RVs remaining available for their ADS function.
The Main Steam system is required by TS 3.5.1 to have six operating S/RVs for ADS. TS Bases 3.5.1 states that each of the ADS S/RVs must be equipped with one air accumulator and associated inlet check valves, in order to provide pneumatic actuation. TS Bases 3.5.1 cites NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," which evaluated the effect of one ADS valve being out of service. Based on this analysis, five operable ADS valves will provide the required depressurization. However, the overall reliability of the ADS system is reduced, because any further valve failure could result in a reduction of depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day completion time was based on a reliability study cited in "Recommended Interim Revisions to LCOs for ECCS Components" and has been found to be acceptable through operating experience.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event:
The ADS system consists of six S/RVs as a defense-in-depth measure; only four of the six ADS valves are required for system operability. The five remaining S/RVs were unaffected by this event and were operable throughout the duration of this event. MSL B Relief Valve actuation comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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 through the ADS or MSRV Automatic Actuation Logics were not affected under normal power.
The valve's mechanical setpoints were unaffected by this event, and they remained capable of lifting.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shut down 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 the reactor was shutdown.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from discovery of the failure until the train was returned to service:
While one ADS valve was inoperable for longer than permitted by TS, the ADS system itself remained operable throughout the duration of this event. Approximately one day elapsed between the time of discovery and returning the valve to service.
VI. Corrective Actions:
Corrective Actions are being managed by TVA's corrective action program under Condition Report (CR) 1161911.
A. Immediate Corrective Actions
A troubleshooting work order was initiated, which found a stab on the back of the MCC breaker cubicle that required adjustment. After adjusting the MCC bucket stab, the green light operated as designed.
B. Corrective Actions to Prevent Recurrence
Corrective actions to prevent recurrence include determining the scope of MCC breakers that feed loads that have a normal and alternate power source. Preventative Maintenance Change Requests were initiated for these MCC Breakers to revise their maintenance PMs to verify PMT includes power source isolation prior to closing the breaker under load. Additionally, a work order will be generated to replace the breaker bus stabs. Finally, a Performance Analysis and Training Needs Analysis will be prepared to develop targeted needs that will close the apparent knowledge gap associated with the implementation of breaker PMT.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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.
3. LER NUMBER
2016 - 00
VII. Additional Information:
A. Previous Similar Events:
There were no previous similar events.
B. Additional Information:
There is no additional information.
C. Safety System Functional Failure Consideration:
In accordance with NUREG-1022, this event is not considered a safety system functional failure.
While one ADS valve was inoperable for longer than permitted by TS, the remaining five ADS valves remained operable throughout this event.
D. Scram with Complications Consideration:
This event did not result in a reactor scram.
VIII. COMMITMENTS
There are no new commitments.
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