07-07-2017 | On May 8, 2017, the Tennessee Valley Authority was presented with as-found testing results indicating that four of the thirteen Main Steam Relief Valves ( MSRVs) from Browns Ferry Nuclear Plant, Unit 2, were outside the +/- 3 percent setpoint band required for their operability. Troubleshooting determined that three MSRVs exceeded their setpoints when their valve discs failed by corrosion bonding to their valve seats. The valve discs were previously platinum coated to prevent this, but the valve seat's rough Stellite surface caused the coating to delaminate. This was the first Unit 2 MSRV service interval to implement the improved surface treatment since a resolution to the delamination issue was identified in 2015. The valve which failed below its setpoint band was determined to have a faulty pilot spring.
These four MSRVs were found to have been inoperable for an indeterminate period of time between April 9, 2015, and February 25, 2017, and longer than permitted by Technical Specifications. The affected valves remained capable of maintaining reactor pressure within American Society of Mechanical Engineers code limits. Additionally, the valves' ability to open under remote-manual operation, activation through the Automatic Depressurization System, or MSRV Automatic Actuation Logics were not affected. The valves remained capable of performing their required safety function.
Corrective Actions were to replace all thirteen Unit 2 MSRV pilot valves with pilot valves which had the platinum coating applied in accordance with the revised procedure, and to analyze the pilot valves of the inoperable MSRVs.
The pilot spring was replaced inside the valve which failed below its specification. |
---|
|
---|
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 100 percent power.
II. Description of Event
A. Event Summary On May 8, 2017, National Technical Systems (NTS) Laboratories provided Tennessee Valley Authority (TVA) with the as-found testing results of the thirteen Main Steam [SB] Relief Valves (MSRVs) [RV] which were removed during the Spring 2017 Unit 2 Refueling Outage 19 (U2R19).
Four of the Main Steam Line B, C, and D Relief Valves (BFN-2-PCV-001-0022, BFN-2-PCV-001-0023, BFN-2-PCV-001-0030, and BFN-2-PCV-001-0041) had as-found lift settings which were outside of the +/- 3 percent band of their setpoints required for their operability.
Technical Specification (TS) 3.4.3 requires twelve of the thirteen Safety/Relief Valves (S/RVs) to be operable for S/RV system operability. These four MSRVs were found to have been inoperable for an indeterminate period of time between April 9, 2015, and February 25, 2017, and longer than permitted by TS 3.4.3.
MSRV operability was restored on April 1, 2017, upon completion of post-maintenance testing (PMT) following the biennial scheduled replacement of the MSRV pilot valves with refurbished valves which were certified to lift within +/- 1 percent of their setpoints.
Throughout this event, the two-stage MSRV pilot valves remained capable of maintaining the reactor pressure below 1375 psig, which is the American Society of Mechanical Engineers (ASME) code limit of 110 percent of the vessel design pressure. The valves remained capable of performing their required safety function.
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.
2017 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 April 9, 2015 February 25, 2017 April 1, 2017 May 8, 2017 Occurrence Unit 2 entered Mode 2, beginning the cycle.
BFN, Unit 2, entered Mode 4, beginning the U2R19.
Replacement of MSRV pilot valves and their associated PMT is completed, as part of the U2R19.
NTS Laboratories provided TVA with the as-found testing results of the thirteen Unit 2 MSRV pilot valves removed during the U2R19.
D. Manufacturer and model number of each component that failed during the event The failed components were all Target Rock Corporation two-stage pressure control valves, model number 7567F.
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 Failure was discovered at NTS Laboratories, during their as-found testing of the thirteen MSRV two-stage pilot valves which were removed during the U2R19.
G. The failure mode, mechanism, and effect of each failed component Three of the four two-stage pilot valves failed due to the corrosion bonding of the valve disc to the valve seat. The remaining pilot valve failed to meet its setpoint band requirements due to a weak spring.
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.
2017 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.
Browns Ferry Nuclear Plant, Unit 2 05000-260 - 004 - 00
III. Cause of the event
A. Cause of each component or system failure or personnel error Three two-stage pilot valves failed above their setpoint bands due to their valve discs corrosion bonding to the valve seat, which is apparently caused by a valve disc surface finish that does not make allowance for corrosion bonding.
One two-stage pilot valve failed below its setpoint band due to a faulty valve spring.
B. Cause(s) and circumstances for each human performance related root cause No human performance related root causes were identified.
IV. Analysis of the event
The 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 MSRV pilot valve inoperability resulted from their setpoints gradually drifting during the course of their operating cycle, which began when Unit 2 entered Mode 2 on April 9, 2015, and lasted until February 25, 2017, when Unit 2 entered Mode 3.
BFN, Unit 2, TS Limiting Condition for Operation (LCO) 3.4.3 requires twelve Operable S/RVs during Modes 1, 2, and 3. If one or more required S/RVs becomes inoperable, Required Action A.1 requires BFN, Unit 2, to enter Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Required Action A.2 requires entering Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. S/RV Operability is defined as being within a +/- 3 percent band of their setpoint values, in accordance with Surveillance Requirement 3.4.3.1. BFN, Unit 2, has thirteen MSRVs to satisfy this requirement with margin.
After the installation of the S/RVs, the as-left lift setpoint for S/RVs BFN-2-PCV-001-0022 and BFN-2-PCV-001-0030 was 1145 psig, S/RV BFN-2-PCV-001-0023 was 1135 psig, and SRN BFN-2-PCV-001-0041 was 1155 psig. However, during as-found testing at NTS Laboratories, the Main Steam Line B Relief Valve, BFN-2-PCV-001-0022, lifted at 1179.5 psig (+3.01 percent); the Main Steam Line C Relief Valves, BFN-2-PCV-001-0023 and BFN-2-PCV-001-0030, lifted at 1183 psig (+4.22 percent) and 1105 psig (-5.76 percent), respectively; and the Main Steam Line D Relief Valve, BFN-2-PCV-001-0041, lifted at 1198 psig (+3.72 percent). During the fuel cycle, these valve lift setpoints gradually drifted outside of the +/- 3 percent margin which is required for their operability. The valves which exceeded their setpoint bands failed due to corrosion bonding between the valve discs and their seats. The valve which failed below its setpoint band was determined to entered Mode 4 on February 25, 2017. Therefore, BFN, Unit 2, operated with inoperable S/RVs for longer than allowed by TS.
2017 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 NEOB-10202, (3150-0104). Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Prior to startup from the Unit 2 Refueling Outage 18, all thirteen BFN, Un't 2 MSRV pilot valves were replaced with refurbished valves which were certified to lift within +/- 1 percent of their setpoint.
Operating Experience (OE) has shown that Target Rock two-stage MSRV setpoint drift is not a uniform, linear process. The corrosion bonding increases at a random rate, and the faulty pilot spring gradually lost its tension over its lifetime. Without an accurate and reliable model for predicting or estimating the setpoint drift development, the point in time where the setpoint exceeded the +/- 3 percent limit cannot be reliably determined. Since this drift occurred between the beginning of the fuel cycle on April 9, 2015, and February 25, 2017, when Unit 2 entered Mode 3, the MSRV inoperability was conservatively declared to have been inoperable for an indeterminate period of time between April 9, 2015, and February 25, 2017, and longer than permitted by plant TS 3.4.3.
V. Assessment of Safety Consequences
System availability was not impacted by this event. The failure of the Automatic Depressurization System (ADS) valves, BFN-2-PCV-001-0022 and BFN-2-PCV-001-0030, to meet the mechanical setpoint identified in TS 3.4.3 does not impact its remote-manual operation, or activation through the ADS or MSRV Automatic Actuation Logics since these operating modes and functions rely upon an electrical signal to energize the MSRV control air solenoid which electrically opens the pilot valve.
TS Bases 3.4.3 states that the overpressure protection system must accommodate the most severe pressurization transient. The MSRVs remained capable of maintaining the reactor pressure below 1375 psig, which is the ASME code limit (110 percent of the vessel design pressure). The valves remained capable of performing their required safety function.
The bounding maximum over-pressurization analyses are performed each fuel cycle to show that the requirements of the ASME code regarding overpressure protection are met. The analyses are performed specifically to show that the dome pressure TS limit of 1325 psig is not exceeded and that the vessel pressure does not exceed the limit of 1375 psig. In addition, the Anticipated Transient Without Scram (ATVVS) pressurization analyses are also performed to demonstrate that the 1500 psig peak vessel pressure limit is not exceeded.
NUREG-0800 defines bases the acceptance criteria for reactor coolant pressure on the ASME Service Level C limits, which are approximately 10.3 MPa (1500 psig) for Boiling Water Reactors (BWRs). For the ATWS analysis, the setpoint groupings conservatively bound the ten lowest as- found MSRV opening setpoints; however, the three highest valve setpoints fall outside the bounds of the valve groupings. Therefore, the limiting ATWS overpressurization event, identified as the ATWS pressure regulator failed open at 100% rated power and 81% rated flow at the beginning of cycle exposure, was re-analyzed by AREVA. The results from this analysis indicates that despite that four MSRV pilot valves were outside their as-found setpoint band, the maximum vessel pressure and maximum dome pressure reaches a maximum of 1398 psig, and therefore does not exceed the ATWS vessel pressure limit of 1500 psig. Therefore, there was no impact on the MSRVs to perform their specified safety function.
- 004 2017 - 00 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.
Logic) installed, which provides defense-in-depth against MSRV setpoint drift by electrically opening MSRV groups based upon setpoints at 1135 psig, 1145 psig, and 1155 psig. Therefore, during a reactor pressure transient event, the four 1135 psig group MSRVs, followed by the four 1145 psig group MSRVs, and finally the five 1155 psig group MSRVs would receive an electrical open signal, providing a defense-in-depth function to allow the valves to perform their safety function.
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 the 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 TS 3.4.3 requires twelve of the thirteen S/RVs to be operable for S/RV system operability. The four failed MSRV pilot valves rendered the entire S/RV system inoperable for the duration of the fuel cycle, from April 9, 2015, to February 25, 2017.
VI. Corrective Actions
Corrective Actions are being managed by TVA's corrective action program under Condition Reports (CRs) 962223 and 1294336.
A. Immediate Corrective Actions
All thirteen of the Unit 2 MSRV pilot valves were replaced with refurbished valves during the U2R19. These valves had platinum coatings applied in accordance with the revised procedure.
As-left testing verified that these refurbished pilot valves were within +/- 1 percent of their name plate setpoints.
2017 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.
Mean Square (RMS) optical comparator to verify that the pilot disc finish quality is 32 RMS or better, optimally 16 RMS, prior to platinum coating.
The pilot spring was replaced inside the valve which failed below its specification.
VII. Previous Similar Events at the Same Site A search of BFN LERs for Units 1, 2, and 3, identified nine LERs for this same issue within the last nine years.
A search of the Corrective Action Program for BFN, Units 1, 2, and 3, identified eighteen MSRV failure events since 1999. These failures were captured by CRs 37328, 59786, 50084, 61823, 81376, 102298, 124944, 146189, 175990, 159200, 226627, 294506, 372047, 558488, 962223, 1157981, 1237184, and 1294336. These individual failures were collectively evaluated and addressed by CR 112190.
CR 55557 identified that corrosion bonding between the valve discs and their seat surfaces were strong enough to cause significant drifts in their opening pressure and reset setpoints. As a corrective action, a design change was implemented to update/install a safety related pressure switch logic that was endorsed by the BWR Owners Group to resolve setpoint drift issues.
CR 56793 identifies the corrosion-prone materials used in the construction of Target Rock two-stage S/RVs presented a fundamental design deficiency. An alternate MSRV design/logic/manufacturer was sought, but industry OE demonstrated that the use of platinum coated valve discs provided the best results.
The Corrective Actions for CR 146189 required platinum coated MSRV discs to be installed in future outages to prevent recurrence. CR 166147 verified that these platinum coatings were in place.
VIII. Additional Information
There is no additional information.
IX. Commitments There are no new commitments.
2017
|
---|
|
|
| | Reporting criterion |
---|
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 | |
|