05000260/LER-2023-001, Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line
| ML23109A362 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/19/2023 |
| From: | Sivaraman M Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| LER 2023-001-00 | |
| Download: ML23109A362 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 2602023001R00 - NRC Website | |
text
Post Office Box 2000, Decatur, Alabama 35609-2000
April 19, 2023 10 CFR 50.73
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Browns Ferry Nuclear Plant, Unit 2 Renewed Facility Operating License No. DPR-52 NRC Docket No. 50-260
Subject: Licensee Event Report 50-260/2023-001-00
Reference: Non-Emergency Event Notification 56371 - Degraded Condition
The enclosed Licensee Event Report provides details of a pressure boundary leak on Browns Ferry Nuclear Plant, Unit 2. The Tennessee Valley Authority 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 plants Technical Specifications; and 10 CFR 50.73(a)(2)(ii)(A), as any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
There are no new regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Christopher L. Vaughn, Nuclear Site Licensing Manager, at (256) 729-2636.
Respectfully, py,
Manu Sivaraman Site Vice President
Enclosure: Licensee Event Report 50-260/2023-001 Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line U.S. Nuclear Regulatory Commission Page 2 April 19, 2023
cc (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant
ENCLOSURE
Browns Ferry Nuclear Plant Unit 2
Licensee Event Report 50-260/2023-001-00
Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line
See Enclosed
Abstract
On February 17, 2023, during a drywell entry for leak identification, BFN personnel discovered a cracked weld on the 2A Recirculation Pump discharge isolation valve drain line. On February 18, 2023, at 0439, it was determined that this drain line was classified as Americ an Society of Mechanical Engineers Code Class I piping and constitutes part of the BFN, Unit 2 Reactor Coolant System pressure boundary. Operations personnel maintained BFN, Unit 2 in Mode 4 or 5 until the leak was repaired. On February 18, 2023, at 1025 CST, eight-hour Event Notification 56371 was made to the NRC. On March 19, 2023, work was completed to remove and replace the drain line.
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
I. Plant Operating Conditions before the Event
At the time of discovery, Browns Ferry Nuclear Plant (BFN) Unit 2 was in Mode 4 at approximately zero percent power.
II. Description of Event
A. Event Summary
On February 17, 2023, during a drywell entry for leak identification, BFN personnel discovered a cracked weld on the 2A Recirculation Pump discharge isolation valve drain line. On February 18, 2023, at 0439, it was determined that this drain line was classified as American Society of Mechanical Engineers (ASME) Code Class I piping and constitutes part of the BFN, Unit 2 Reactor Coolant System pressure boundary. Operations personnel maintained BFN, Unit 2 in Mode 4 or 5 until the leak was repaired. On February 18, 2023, at 1025 CST, eight-hour Event Notification 56371 was made to the NRC. On March 19, 2023, work was completed to remove and replace the drain line.
The 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 plants Technical Specifications (TS); and 10 CFR 50.73(a)(2)(ii)(A), as any event or condition that resulted in the condition of the nu clear power plant, including its principal safety barriers, being seriously degraded.
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.
C. Dates and approximate times of occurrences
Date Time (CST) Occurrence February 18, 2023 0439 Determined that leaking piping was ASME Code Class I.
February 18, 2023 1025 Made eight-hour EN 56371 to NRC.
March 14, 2023 1611 Completed WO 123487635 to repair leaking drain line.
March 19, 2023 2342 Declared BFN Unit 2 RCS structural integrity operable
D. Manufacturer and model number of each component that failed during the event
There were no failed components related to this event.
E. Other systems or secondary functions affected
No other systems or secondary functions were affected.
F. Method of discovery of each component or system failure or procedural error
During the planned outage on BFN Unit 2, personnel entered the drywell for leak identification and discovered a cracked weld on the 2A Recirc pump discharge isolative valve drain line. An engineering evaluation determined this line to be ASME Code Class I Piping.
G. The failure mode, mechanism, and effect of each failed component
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
H. Operator actions
There were no operator actions related to this event.
I. Automatically and manually initiated safety system responses
There were no automatic or manual safety system responses during this event.
III. Cause of the event
A. Cause of each component or system failure or personnel error
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
IV. Analysis of the event
The Radioactive Material Barrier is the systems, structures, or equipment that, together, physically prevent the uncontrolled release of radioactive ma terials. One of these is the Nuclear System Process Barrier, made up of a Primary Barrier and a Secondary Barrier, which includes the systems of vessels, pipes, pumps, tubes, and simila r process equipment that contain the steam, water, gases, and radioactive materials coming from, going to, or in communication with the reactor core. The Primary Barrier, also know n as the Reactor Coolant Pressure Boundary, consists of the reactor vessel and attached piping out to and including the second isolation valve in each attached pipe. A leak in the Pressure Boundary is therefore a degradation of one of the nuclear power plants principal safety barriers.
BFN, Unit 2 TS Limiting Condition for Operation (LCO) 3.4.4, RCS Operational Leakage, requires that there be no pressure boundary leakage while in Modes 1, 2, or 3. TS LCO 3.4.4 Condition C requires, when any pressure boundary leakage exists, that the unit be in Mode 3 within twelve hours, and Mode 4 within thirty-six hours. A Past Operability Evaluation (POE) determined that this condition is likely to have existed since No vember 13, 2022, when an increase in unidentified DW leakage from the RCS was detected. Pressure boundary leakage likely existed from this time until February 17, 2023, when BFN, Unit 2 was placed in Mode 4.
V. Assessment of Safety Consequences
The POE for this event determined that the safety significance was low because, per Calculation MDQ0999950033 Rev. 3, Exclusion Criteria for ISI Scope, the normal reactor coolant makeup systems, Reactor Core Isolation Cooling, Feedwater, and Control Rod Drive, have adequate capacity to makeup reactor coolant in the worst-case scenario of a small break Loss of Coolant Accident (LOCA). Additionally, Calculation NDQ0074880118 Rev.8, Evaluation of LPCI Flow to RPV with Failed Open Min Flow Bypass Valve shows that the LPCI mode of the RHR system can still meet the required LPCI flow to the RPV in the worse-case scenario of a small break LOCA for the duration of its mission time.
During this time the HPCI system [BJ], both loops of the Core Spray system (CS) [BG], and all ADS valves remained operable when required, asi de from periods of planned maintenance and testing, and were available to provide coolant flow to the core during an emergency.
Based on the above, the TVA has concluded that sufficient systems were available to provide the required safety functions needed to protect the health and safety of the public.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event
There were no systems or components other than the RCS Pressure Boundary that failed during the event.
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
During the period concurrent with this event where BFN, Unit 2 was in Mode 4 or Mode 5, all systems or components required 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 in Mode 4 or Mode 5 remained operable, aside from periods of planned maintenance and testing.
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
This event did not result in the inoperability of any safety system train.
VI. Corrective Actions
Corrective Actions are being managed by the TVAs corrective action program under Condition Report (CR) 1836572.
A. Immediate Corrective Actions
The immediate corrective action for this event was to repair the valves drain line under Work Order 123487635.
B. Corrective Actions to Prevent Recurrence or to reduce the probability of similar events occurring in the future
There will be a supplement to this Licensee Event Report to provide additional time to obtain failure analysis results for the drain line associated with 2-FCV-68-3.
VII. Previous Similar Events at the Same Site
There are two similar events which have occurred at BFN within the last two years:
- 1. On January 15, 2022, at 2320 CST, during a drywell entry for leak identification, BFN Engineering personnel discovered a through-wall piping leak on a test line upstream of the RHR and SDC test shut-off valve. This test li ne is classified as ASME Code Class 1 piping and constitutes part of the BFN, Unit 1 RCS pressure boundary.
The root cause of this event was small bore piping which was not specifically analyzed for fatigue failure vulnerability due to operational or resonance vibration. The corrective action for this event is to implement Engineering Change Packages for all small bore piping with vulnerability to fatigue failure due to exceeding the endurance limit due to operational vibration.
This event was reported to the NRC under Licensee Event Report (LER) 50-259/2022-001 Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line.
- 2. On December 2, 2022, at 2330 CST, during a drywell entry for leak identification, BFN Engineering personnel discovered a through-wall piping leak on a test line between the two RHR Shutdown Cooling test line isolation valves. On December 3, 2022, at 1000, it was determined that this test line was classified as ASME Code Class 1 piping and constitutes part of the BFN, Unit 3 RCS pressure boundary.
The root cause of the event has not yet been determined.
This event was reported to the NRC under LER 50-296/2022-003 Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line
VIII. Additional Information
There is no additional information.
IX. Commitments
There are no new commitments.