10-31-2017 | On September 1, 2017, at approximately 1006 Central Daylight Time ( CDT), Browns Ferry Nuclear Plant ( BFN) Unit 3 3A Residual Heat Removal ( RHR) system pump failed to start during performance of Surveillance 3-SR-3.5.1.6 ( RHR I), Quarterly RHR System Rated Flow Test Loop I. The apparent cause was the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance was revised to include steps to secure the breaker's mounting hardware which caused internal binding of the indication flag. Binding of the indication flag prevented the closing spring of the breaker from charging and the breaker from closing on demand. As a result, automatic start of the 3A RHR pump was prevented. On September 1, 2017, at approximately 1633 CDT, the 3A RHR Pump was declared operable following lubrication and testing of the breaker's indication flag mounting bolt.
A Past Operability Evaluation concluded that the 3A RHR Pump was inoperable from July 26, 2017 to September 1, 2017, which exceeded the Technical Specification allowed outage time. During this time, the 3B, 3C, and 3D RHR pumps would have started automatically upon receipt of an Emergency Core Cooling System ( ECCS) initiation signal or from an Operator manual start demand from the Control Room. Based on results from the Probability Risk Assessment and Engineering inspections, there was no significant risk to the health and safety of the public or plant personnel for this event. The Corrective Action to reduce the probability of similar events occurring in the future will be addressed by revising the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens breakers to ensure freedom of movement of the indication flag is present during the breaker inspection. |
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Category:Letter
MONTHYEARIR 05000259/20244042024-09-0303 September 2024 Cyber Security Inspection Report 05000259/2024404 and 05000260-2024404 and 05000296/2024404-Cover Letter IR 05000259/20240052024-08-26026 August 2024 Updated Inspection Plan for Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Report 05000259/2024005, 05000260/2024005 and 05000296/2024005 IR 05000259/20244022024-08-0606 August 2024 Security Baseline Inspection Report 05000259/2024402 and 05000260/2024402 and 05000296/2024402 ML24219A0272024-08-0606 August 2024 Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations IR 05000259/20240022024-08-0202 August 2024 Brown Ferry Nuclear Plant – Integrated Inspection Report05000259/2024002 and 05000260/2024002 and 05000296/2024002 ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24183A4142024-07-10010 July 2024 – License Renewal Regulatory Limited Scope Audit Regarding the Environmental Review of the License Renewal Application (EPID Number: L-2024-SLE-0000) (Docket Numbers: 50-259, 50-260, and 50-296) ML24184A1142024-07-0202 July 2024 Site Emergency Plan Implementing Procedure Revision ML24183A3842024-07-0101 July 2024 Registration of Use of Cask to Store Spent Fuel (MPC-364, -365) ML24179A0282024-06-26026 June 2024 Evaluation of Effects of Out-of-Limits Condition as Described in IWB-3720(a) ML24175A0042024-06-23023 June 2024 Interim Report of a Deviation or Failure to Comply Associated with a Valve in the Unit 3 High Pressure Coolant Injection System ML24176A1132024-06-23023 June 2024 American Society of Mechanical Engineers, Section XI, Fourth 10 Year Inspection Interval, Inservice Inspection, System Pressure Test, Containment Inspection, and Repair and Replacement Programs, Owner’S Activity Report Cycle 21 Oper ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24155A0042024-06-18018 June 2024 Proposed Alternative to the Requirements of the ASME Code (Revised Alternative Request 0-ISI-47) ML24158A5312024-06-0606 June 2024 Registration of Use of Cask to Store Spent Fuel (MPC-361, -362, -363) ML24071A0292024-06-0505 June 2024 Subsequent License Renewal Application Enclosure 3 - Proprietary Determination Letter ML24068A2612024-06-0505 June 2024 SLRA Fluence Methodology Report - Proprietary Determination Letter IR 05000259/20244032024-05-22022 May 2024 – Security Baseline Report 05000259/2024403 and 05000260/2024403 and 05000296/2024403 ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc ML24136A0702024-05-15015 May 2024 2023 Annual Radiological Environmental Operating Report IR 05000259/20240012024-05-14014 May 2024 Integrated Inspection Report 05000259/2024001, 05000260/2024001, and 05000296/2024001 CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24123A2012024-05-0202 May 2024 NRC Cybersecurity Baseline Inspection (NRC Inspection Report 05000259/2024404, 05000260-2024404, 05000296/2024404) and Request for Information ML24122A6852024-05-0101 May 2024 2023 Annual Radioactive Effluent Release Report and Offsite Dose Calculation Manual CNL-24-036, – 10 CFR 50.46 Annual Report2024-04-25025 April 2024 – 10 CFR 50.46 Annual Report ML24116A2522024-04-25025 April 2024 Site Emergency Plan Implementing Procedure Revision CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 ML24087A2302024-04-18018 April 2024 Exemption from Select Requirements of 10 CFR Part 73, Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24106A2492024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Hoskin, Jr., Cherokee Nation ML24102A0392024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Johnson, Absentee Tribe of Shawnee ML24106A2572024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Yahola, Kialegee Tribal Town ML24106A2432024-04-16016 April 2024 Letter to Federally Recognized Tribes-Ben, Mississippi Band of Choctaw Indians ML24106A2512024-04-16016 April 2024 Letter to Federally Recognized Tribes-Morrow, Thlopthlocco Tribal Town ML24107A0022024-04-16016 April 2024 Letter to Federally Recognized Tribes-Bunch, United Keetoowah Band of Cherokee Indians ML24106A2402024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Barnes, Shawnee Tribe ML24106A2632024-04-16016 April 2024 Ltr to State Recognized Tribes-Hamilton, Cher-O-Creek Intra Tribal Indians ML24102A0422024-04-16016 April 2024 Ltr to State Recognized Tribes-Russell, Cherokee Tribe of Northeast Alabama ML24106A2542024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Rogers, Jena Band of Choctaw Indians ML24106A2622024-04-16016 April 2024 Ltr to State Recognized Tribes-Gilmore, Southeastern Mvskoke Nation ML24106A2582024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Yargee, Alabama-Quassarte Tribal Town ML24106A2382024-04-16016 April 2024 Letter to Federally Recognized Tribes-Anoatubby, Chickasaw Nation ML24106A2452024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Cernek, Couchatta Tribe of Louisiana ML24106A2472024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Hicks, Eastern Band of Cherokee Indians ML24106A2462024-04-16016 April 2024 Letter to Federally Recognized Tribes-Cypress, Miccosukee Tribe ML24106A2592024-04-16016 April 2024 Ltr to State Recognized Tribes-Bridges, United Cherokee Aniynwiya ML24106A2482024-04-16016 April 2024 Letter to Federally Recognized Tribes-Hill, Muscogee (Creek) Nation ML24106A2422024-04-16016 April 2024 Ltr to Federally Recognized Tribes-Batton, Choctaw Nation of Oklahoma 2024-09-03
[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 3 was in Mode 1 at 100 percent power.
II. Description of Event
A. Event Summary B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event No inoperable systems, structures, or components contributed to this event.
On September 1, 2017, at approximately 1006 Central Daylight Time (CDT), BFN Unit 3 3A Residual Heat Removal (RHR)[BO] system pump [P] failed to start during the performance of Surveillance, 3-SR-3.5.1.6 (RHR I), Quarterly RHR System Rated Flow Test Loop I. Maintenance troubleshooting revealed that the 3A RHR pump motor breaker's [BKR] closing spring failed to charge preventing the breaker from closing on demand. Malfunction of the 3A RHR pump motor breaker resulted in 3A RHR pump failing to start during manual or automatic actuation. Operations personnel declared 3A RHR pump inoperable.
A Past Operability Evaluation (POE) was performed for the 3A RHR pump failure. The POE concluded that the 3A RHR pump was inoperable from July 26, 2017 to September 1, 2017. BFN Technical Specifications (TS) Limiting Condition of Operability (LCO) 3.5.1 Required Action A.1 requires, with one required Emergency Core Cooling System (ECCS) injection/spray subsystem inoperable, that the required ECCS injection/ spray subsystem be restored to operable status within seven days. Additionally, BFN TS LCO 3.6.2.3 Required Action A.1 requires, with one RHR suppression pool cooling subsystem inoperable, restore the RHR suppression pool cooling subsystem to operable status within 30 days. The 3A RHR pump was inoperable for a time longer than allowed by TS. Therefore, 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.
- 00 2017 - 001 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
E. Other systems or secondary functions affected
There were no other systems or secondary functions affected by this condition.
F. Method of discovery of each component or system failure or procedural error G. The failure mode, mechanism, and effect of each failed component D. Manufacturer and model number of each component that failed during the event The failed component during this event was the 3A RHR pump motor breaker. The model number for the 3A RHR pump motor breaker is 5-3AF-GEH-250-1200-58, manufactured by Siemens.
It was discovered that although the 3A RHR pump motor breaker indicated charged and ready to close, the breaker actuating spring was actually discharged. Additional troubleshooting revealed that the breaker's charged indicating flag was bound, preventing actuation of the internal breaker On September 1, 2017, during performance of Surveillance 3-SR-3.5.1.6 (RHR I), 3A RHR pump failed to start. Maintenance troubleshooting determined that the closing spring on 3A RHR pump motor breaker failed to charge due to internal binding of the breaker's indicating flag. The as-found condition of the breaker prevented the breaker from closing on demand. Due to the malfunction of 3A RHR pump motor breaker, the 3A RHR pump was unable to perform its design basis function.
This event was the result of internal binding of the 3A RHR pump motor breaker's charged/uncharged indication flag, which prevented the closing spring from charging and the breaker from closing on demand. With the breaker's actuating spring discharged, manual and automatic start of the 3A RHR pump was prevented.
July 26, 2017, 0859 CDT 3A RHR pump was started in Suppression Pool cooling mode to support Reactor Core Isolation Cooling (RCIC) [BN] flowrate surveillance testing.
July 26, 2017, 1110 CDT 3A RHR pump was removed from service following RCIC flowrate testing.
September 1, 2017, 1006 CDT During performance of Surveillance 3-SR-3.5.1.6 (RHR I), 3A RHR pump failed to start. Operations personnel declared 3A RHR pump inoperable.
September 1, 2017, 1633 CDT Operations personnel placed 3A RHR pump in service following Maintenance troubleshooting and post maintenance testing (PMT).
Dates & Approximate Times Occurrence - 00 2017 - 001 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.
H. Operator actions
In response to indication that the 3A RHR pump was not functional, Operations personnel declared 3A RHR pump inoperable.
I. Automatically and manually initiated safety system responses
No safety system responses resulted from this event.
III. Cause of the event
A. Cause of each component or system failure or personnel error charging limit switches. Since the limit switches could not change position following the last operation of the breaker, the ready to close light never extinguished and the charging motor never charged the breaker's actuating spring. This failure mechanism resulted in a false indication of the status of the breaker.
The Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance was revised on September 30, 2010 to include steps to verify that the breaker's spring charging motor mounting screws were secure. The procedure did not specify a step to inspect the indication flag mounting bolts. The mounting bolts are installed and inspected at the vendor's facility and are not normally adjusted during Preventive Maintenance (PM). Therefore, torque requirements were never provided for the indication mounting bolts. The inclusion of steps in the procedure to secure the spring charging motor mounting screws, without providing torque requirements, could have inadvertently caused overtightening of the indication flag. Therefore, any Wyle/Siemens breaker tested after September 30, 2010, would be more susceptible to binding of the indication flag. Thus, the apparent cause was the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance provided a potential inaccurate step to secure all mounting hardware.
A 20 year old breaker test specimen was inspected at the vendor's facility. The vendor simulated the same failure mode of the breaker by tightening the indication flag mounting screw. When the mounting screw was tightened, it caused the indication flag to bind and become resistant to movement. However, loosening of the mounting screw allowed freedom of range of movement and the indication flag operated as designed.
The direct cause of this event was binding of the charged/uncharged indication flag, which prevented the closing spring from charging and the 3A RHR pump motor breaker from closing on demand. Malfunction of the breaker resulted in the 3A RHR pump failing to start during manual or automatic actuation.
- 00 2017 - 001 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.
IV. Analysis of the event
A POE was performed for the 3A RHR pump. The POE concluded that the 3A RHR pump was inoperable from July 26, 2017 to September 1, 2017. BFN, Unit 3, TS LCO 3.5.1 requires that each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be Operable in Mode 1, and Modes 2 and 3, except High Pressure Coolant Injection (HPCI) [BJ] and ADS valves are not required to be operable with reactor steam dome pressure less than or equal to 150 pounds per square inch gauge (psig). Required Action A.1 requires, with one required ECCS injection/spray subsystem inoperable, that the required ECCS injection/ spray subsystem be restored to operable status within seven days. Additionally, BFN TS LCO 3.6.2.3 requires that four RHR suppression pool cooling subsystems shall be operable. Required Action A.1 requires, with one RHR suppression pool cooling subsystem inoperable, restore the RHR suppression pool cooling subsystem to operable status within 30 days. The 3A RHR Pump, credited as a required low This event resulted in BFN, Unit 3, 3A RHR Pump being inoperable for longer than allowed by the TS.
With the 3A RHR pump motor breaker actuating spring discharged, the ability of the 3A RHR pump to automatically start or manually start was prevented.
The Safety objectives of the RHR System are as follows:
The Low Pressure Coolant Injection (LPCI) subsystem is an integral part of the RHR System. It operates to restore and, if necessary, maintain the coolant inventory in the reactor vessel after a loss-of-coolant accident so that the core is sufficiently cooled. The LPCI mode of operation of the RHR System pumps water into the reactor vessel in time to flood the core and limit fuel clad temperature.
During the period when 3A RHR pump was inoperable, 3B, 3C, and 3D RHR pumps were operable and available to perform this function.
B. Cause(s) and circumstances for each human performance related root cause There were no human performance root causes related to this event. All vendor guidance and maintenance procedures were followed. Although the apparent cause was an inaccurate step in the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance to secure all mounting hardware, it was not reasonably foreseeable that slight tightening of the mounting bolts would cause binding of the indication flag. This step was included to preclude breaker inoperability due to loose bolts.
a. To restore and maintain the coolant inventory in the reactor vessel so that the core is adequately cooled after a loss-of-coolant accident. The RHR System also provides cooling for the pressure suppression pool so that condensation of the steam resulting from the blowdown due to the design basis loss-of-coolant accident is ensured.
b. To further extend the redundancy of the Core Standby Cooling Systems by providing for containment cooling.
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V. Assessment of Safety Consequences
pressure ECCS Injection/spray and suppression pool cooling subsystem, was inoperable for a time longer than allowed by TS. Therefore, Unit 3, was in violation of TS 3.5.1 and 3.6.2.3 Required Action A.1.
A Probabilistic Risk Assessment (PRA) was performed to evaluate the failure of the 3A RHR pump. The PRA concluded that the risk thresholds for the 3A RHR Pump was GREEN during the time of inoperability. The total change in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) for the failure of the 3A RHR pump from July 26, 2017 to September 1, 2017 was 3.06E-07 and 7.39E-08, respectively. The 1E-6 (CDF) and 1E-7 (LERF) thresholds for green were met.
There are a total of 93 safety related 4kV Wyle/Siemens breakers located in the plant. PMs were performed on 60 of the 93 breakers after the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance was revised to include steps to secure the breaker's mounting hardware. These 60 breakers could be more susceptible to binding of the indication flag. Engineering performed a visual inspection of 43 out of the 60 safety related horizontal Wyle/Siemens breakers. These 43 breakers were chosen because they were in a non-protected status, were in the open position to preclude the potential trip of a closed breaker, and did not present a high safety risk to plant personnel. The inspection verified that the closing spring of the breakers were charged as expected. Work orders have been created to perform more intrusive inspections on all 60 breakers to test the indication flag. The 17 breakers that were excluded from the initial visual inspection will also be inspected to verify the closing springs are charged. These inspections will occur during planned outages to mitigate personnel safety risks and the potential of a plant trip or transient. While common mode failure in the remaining breakers is not expected, if the inspections result in identifying similar failures, a supplement will be provided to this LER.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event Based on the above discussion, TVA has concluded that, during the time period that 3A RHR Pump was inoperable, there was no significant risk to the health and safety of the public or plant personnel for this event.
During this event, RHR Pumps 3B, 3C, and 3D retained the ability to automatically start upon receipt of an ECCS Initiation signal or from an Operator manual start demand from the Control Room.
Additionally, the 3A, 3B, 3C and 3D Core Spray pumps remained available and operable except during minimal periods where the core spray pumps were inoperable due to Surveillances being performed.
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VI. Corrective Actions
A. Immediate Corrective Actions
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 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 Corrective Actions (CAs) are being managed by TVA's corrective action program under Condition Report (CR) 1334534.
B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future Additional corrective actions include:
Inspections on all safety related Wyle/Siemens breakers will be performed to inspect freedom of movement of the indication flags and to ensure no other common mode failures exist.
A failure analysis of the breaker will be performed to determine cause of the 3A RHR Pump failure.
This event resulted in inoperability of the 3A RHR pump for a time longer than allowed by TS, from the last time the 3A RHR Pump was started in Suppression Pool Cooling Mode to support RCIC Flowrate Surveillance testing on July 26, 2017, until the time of discovery of the condition on September 1, 2017.
The Corrective Action to reduce the probability of similar events occurring in the future will be addressed by revising the Electrical Preventive Maintenance Instruction for 4kV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance to ensure freedom of movement of the indication flag is present during the Breaker inspection. If binding is present, adjustment of the indication flag mounting bolt will be made until freedom of movement is obtained.
The indication flag mounting bolt was inspected, lubricated and tested. Engineering performed initial inspections of 43 risk significant, safety related horizontal Wyle/Siemens breakers to address common mode failure. The closing springs of the 43 breakers inspected were all charged as expected with no indication of common mode failure.
These conditions did not occur during a shutdown.
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Browns Ferry Nuclear Plant, Unit 3 05000-296 A review of the BFN CAP and Licensee Event Reports (LERs) for Units 1, 2, and 3 revealed three similar events over the last four years:
[RLY] which was rendered inoperable for longer than allowed by TS due to improper landing of leads during PM. The event is similar because an improperly connected wire resulted in a long period of inoperability for a safety system. Corrective Actions were to develop and deliver a case study to the Maintenance, Modifications, and Operations departments based on the details of this event.
control room due to a loose fastener. The event is similar because an improperly latched hand switch resulted in a long period of inoperability for a safety system. Corrective Actions for this event were to discipline the individuals responsible, to tighten the loose fastener, and to revise maintenance instructions to reduce the probability of recurrence.
longer than allowed by TS due to failure of the 3A RHR Pump Motor Breaker Transfer Switch (MBTS) to fully latch due to binding. Binding of the MBTS resulted from being installed greater than it's twenty-one year service life with no PM performed. Corrective Actions were to verify similar SB-1 transfer switches are latched in the NORMAL position on BFN, Units 1, 2, and 3, and to create a PM activity to periodically replace GE SB-1 transfer switches similar to the 3A RHR Pump MBTS on a 20 year frequency.
Corrective Actions from these LERs would not have prevented this event.
VIII. Additional Information
There is no additional information.
IX. Commitments There are no new commitments.
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05000296/LER-2017-001 | 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-2017-001 | 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-2017-001 | Signal Timer for 4kV Shutdown Board C Inoperable for Longer Than Allowed by Technical Specifications due to Detached Restraining Strap | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000296/LER-2017-002 | 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 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000260/LER-2017-002 | 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 | | 05000259/LER-2017-002 | 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-2017-003 | 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 | | 05000260/LER-2017-004 | 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 | |
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