IR 05000259/2024003
| ML24303A344 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/04/2024 |
| From: | Louis Mckown NRC/RGN-II/DORS |
| To: | Erb D Tennessee Valley Authority |
| References | |
| IR 2024003 | |
| Download: ML24303A344 (26) | |
Text
SUBJECT:
BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2024003 AND 05000260/2024003 AND 05000296/2024003
Dear Delson Erb:
On September 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On October 17, 2024, the NRC inspectors discussed the results of this inspection with Daniel Komm, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as an NCV consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Browns Ferry Nuclear Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Browns Ferry Nuclear Plant.
November 4, 2024 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Louis J. McKown, II, Chief Projects Branch 5 Division of Operating Reactor Safety Docket Nos. 05000259 and 05000260 and 05000296 License Nos. DPR-33 and DPR-52 and DPR-68
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000259, 05000260 and 05000296 License Numbers:
DPR-33, DPR-52 and DPR-68 Report Numbers:
05000259/2024003, 05000260/2024003 and 05000296/2024003 Enterprise Identifier:
I-2024-003-0019 Licensee:
Tennessee Valley Authority Facility:
Browns Ferry Nuclear Plant Location:
Athens, Alabama Inspection Dates:
July 1, 2024 to September 30, 2024 Inspectors:
S. Billups, Resident Inspector B. Bowker, Senior Reactor Inspector W. Deschaine, Senior Project Engineer J. Diaz-Velez, Senior Health Physicist A. Nielsen, Senior Health Physicist K. Pfeil, Resident Inspector D. Restrepo, Health Physicist A. Ruh, Senior Reactor Inspector T. Steadham, Senior Resident Inspector D. Strickland, Senior Reactor Inspector Approved By:
Louis J. McKown, II, Chief Projects Branch 5 Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Browns Ferry Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. A licensee-identified non-cited violation is documented in report section: 71124.0
List of Findings and Violations
Failure to Maintain Reactor Coolant System Thermal Limit Parameters within Technical Specification Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000259/2024003-01 Open/Closed
[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 1 Technical Specification (TS) 5.4.1.a, "Procedures," was identified when the licensee failed to implement written procedures as specified in Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2, Appendix A,
Section 5. Specifically, following the Unit 1 scram on April 24, 2024, the licensee failed to implement the direction of abnormal operating procedures 1-AOI-100-1 and 1-AOI-68-1 to either restart the recirculation pumps or initiate a plant cooldown to maintain compliance with TS surveillance requirement 3.4.9.1, reactor coolant system pressure and temperature limits.
Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000260/2024003-02 Open/Closed Not Applicable 71153 A self-revealed Severity Level IV non-cited violation of Technical Specifications 3.4.3 and 3.0.4 was identified when the licensee discovered, through as-found test results, that four of thirteen main steam relief valves removed for testing had as-found lift settings outside of the
+/-3 percent setpoint band required for operability.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000260/2023-002-00 Unit 2, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at full (100 percent) rated thermal power (RTP). On July 19, 2024, operators lowered reactor power to 67 percent RTP for a control rod sequence exchange. On July 20, 2024, the unit was returned to 100 percent RTP. On August 2, 2024, operators lowered reactor power to 73 percent RTP for a control rod sequence exchange. On August 3, 2024, the unit was returned to 100 percent RTP. On August 9, 2024, operators lowered reactor power to 73 percent RTP for a control rod sequence exchange. On August 10, 2024, the unit was returned to 100 percent RTP. On August 16, 2024, operators lowered reactor power to 73 percent RTP for a control rod sequence exchange. On August 17, 2024, the unit was returned to 100 percent RTP. On August 30, 2024, operators shut down the unit for a planned refueling outage. Following refueling activities, operators restarted the unit on September 27, 2024. Operators continued power ascension for the remainder of the inspection period where reactor power was at 88 percent RTP.
Unit 2 began the inspection period at full RTP. On August 17, 2024, operators lowered reactor power to 59 percent RTP to clean the main condenser water boxes to restore condenser vacuum margin and to perform a control rod sequence exchange. Following maintenance on August 19, 2024, the unit was restored to 100 percent RTP where it operated at or near for the remainder of the inspection period.
Unit 3 began the inspection period at full RTP. On September 8, 2024, operators lowered reactor power to 60 percent RTP for a control rod sequence exchange. On September 8, 2024, the unit was returned to 100 percent RTP where it operated at or near for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect auxiliary decay heat removal and the switchyard from impending severe weather of the incoming tropical storm with the potential for tornadoes on September 12, 2024.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Main bank 3 and battery board 3 after maintenance and testing on July 3, 2024
- (2) Unit 1 high pressure coolant injection following flow rate test on July 18, 2024
- (3) Unit 2 core spray loop 2 following flow rate test on July 23, 2024
- (4) Unit 3 residual heat removal loop 2 following maintenance on August 22, 2024
- (5) Unit 3 high pressure coolant injection following maintenance on August 23, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 1, Fire Area 26, turbine building elevation 586' on July 10, 2024
- (2) Unit 1, Fire Area 1-1, reactor building, southwest quad elevation 519', 541' and high-pressure coolant injection room on July 18, 2024
- (3) Unit 2, Fire Area 2-2, reactor building, northeast quad elevation 519' and 541' on July 23, 2024
- (4) Unit 1, reactor building steam vault on September 03, 2024
- (5) Unit 1, drywell general areas on September 10, 2024
- (6) Unit common, FLEX equipment storage building on September 16, 2024
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the Unit 1 residual heat removal pump rooms.
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01)
The inspectors evaluated boiling water reactor nondestructive testing by reviewing the following examinations from September 9 - September 12, 2024.
- (1) Ultrasonic Examination
- KMS-1-89, Pipe to Branch Connection, ASME Class 1
- GFW-1-15, Reducing Tee to Elbow, ASME Class 1 Liquid Penetrant Examination
- 1-47B452-3041-IA, RHR Welded Attachment, ASME Class 1
- RWCU-1-005-006, Forged Valve to Cast Valve, ASME Class 1 Visual Examination
- RPV-SUPP-1-1, VT-3, Anchor, ASME Class 1 Magnetic Particle Examination
- RCH-1-2C, Head to Flange Weld, ASME Class 1 Welding Activities
- Work order (WO) 122996828, Install new CKV-1-875 valve to reduce ALT leak boundary, ASME Class 2
- WO 122179650, Modify high pressure coolant injection steam supply pipe support 1-47B455-2134, ASME Class 2
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during Unit 1 startup activities including removing shutdown cooling from service, placing the reactor mode switch to start/hot standby, and control rod withdrawal to criticality on September 27, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (6 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Main battery bank 3 breaker maintenance on July 9, 2024
- (2) Unit common containment air dilution system maintenance on August 21, 2024
- (3) Unit 2 and 3 primary containment integrity and isolation associated with traversing incore probe systems on August 22, 2024
- (4) Unit 1 emergency high pressure makeup system maintenance on August 23, 2024
- (5) Unit 3 reactor core isolation cooling maintenance on September 10, 2024
- (6) Unit 1/2 control bay chiller maintenance on September 26, 2024
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
- (1) Unit 2 high pressure coolant injection rupture disc commercial grade dedication on August 22, 2024
Aging Management (IP Section 03.03) (1 Sample)
The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:
- (1) South power loop electrical feed to auxiliary decay heat removal on September 26, 2024.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Maintenance risk during Unit 1 oil leak on 1A electro-hydraulic control pump discharge header on July 17, 2024
- (2) Emergent work control for Unit 3 relay 211-603 inoperable on August 23, 2024
- (3) Maintenance risk assessment during Unit 1 residual heat removal shutdown cooling outage on September 11, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Condition report (CR) 1942327, Unit 1 east scram discharge volume drain control valve failed to fully open during valve operability test on July 15, 2024
- (2) CR 1952331, Unit 3 phase unbalance relay failure on 4kV shutdown board 3EA on August 23, 2024
- (3) CR 1742531, Unit 2 high pressure coolant injection steam exhaust line pressurized to 160 psig on September 3, 2024
- (4) CR 1955855, Unit 1 control rod drive power supply 1-PX-85-3A/PS6 failed resister on September 5, 2024
- (5) CR 1957254, Unit 1 & 2 airlock door will not allow ingress or egress to or from either unit reactor building on September 10, 2024
- (6) CR 1954797, Unit 1 4kv shutdown board 1 failed to transfer to alternate on September 12, 2024
- (7) CR 1958906, turbine control valve no. 1 fast acting reactor protection system input pressure switch as found setpoint low on September 16, 2024
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) DCN 71458, add supports for common unit residual heat removal service water cables in handhole 15
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated various Unit 1 refueling 1R15 activities from August 30, 2024, through September 26, 2024.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (8 Samples)
- (1) WO 123783014, Unit 2 core spray loop II system venting after maintenance on August 1, 2024
- (2) WO 124804633, Unit 1/2 airlock door magnetic switch reattachment on September 10, 2024
- (3) Final core verification for Unit 1 cycle 15 reload on September 18, 2024
- (4) WO 124808664, replace fuse cut-out for auxiliary decay heat removal electrical service on September 19, 2024
- (5) WO 124804993, Unit 1 residual heat removal valves testing during cold shutdown on September 25, 2024
- (6) WO 123644012, Unit 1 primary containment integrated leak rate test on September 26, 2024
- (7) WO 124154786, Unit 1 high pressure coolant injection surveillance test following extensive system maintenance on September 28, 2024
- (8) WO 124792337, Unit 1 control rod drive power supply 1-PX-85-3A/PS6 resistor replacement on September 30, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) WO 123803867, Unit common maintenance of DC circuit breakers for DC battery board on July 8, 2024
- (2) WO 123794697, Unit 3 standby liquid control pump functional test on July 26, 2024
- (3) WO 123643965, Unit 1 torus spray nozzle test on September 23, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) WO 124543462, Unit 1 residual heat removal pump A periodic pump performance test on September 25, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) WO 123544185, Unit 1 primary containment local leak rate test reactor feedwater line A for penetration X-9A on September 12, 2024
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) WOs 123989262 and 123236511, 4kV combustion turbine generator maintenance quarterly test on September 18,
RADIATION SAFETY
71124.06 - Radioactive Gaseous and Liquid Effluent Treatment
Walkdowns and Observations (IP Section 03.01) (5 Samples)
The inspectors evaluated the following radioactive effluent systems during walkdowns:
- (1) Standby gas treatment system, Train A
- (2) Main stack particulate, iodine, and noble gas weekly sample collection (3)2-RM-90-250, Unit 1 Reactor/Turbine/Refuel Building Ventilation Zone Monitor (4)0-RM-90-130, Liquid Radwaste Effluent Monitor (5)0-RM-90-252, Radwaste Building Vent Monitor
Sampling and Analysis (IP Section 03.02) (3 Samples)
Inspectors evaluated the following effluent samples, sampling processes and compensatory samples:
(1)0-RM-90-306, Wide-range Gaseous Effluent Radiation Monitor, sample line configuration and heat tracing (2)2-RM-90-250, compensatory sampling records, December 4-8, 2021 (3)0-RM-90-147/148, compensatory sampling records, January 25-27, 2023
Dose Calculations (IP Section 03.03) (3 Samples)
The inspectors evaluated the following dose calculations:
- (1) L-20240601-016-C, Unit 1 station sump liquid release permit, July 1, 2024
- (2) G-20240701-530-C, Unit 3 reactor building exhaust gaseous release permit, July 8, 2024
- (3) G-20231017-822-C, main stack gaseous release permit, October 24, 2023
Abnormal Discharges (IP Section 03.04) (1 Sample)
The inspectors evaluated the following abnormal discharges:
- (1) Abnormal liquid release to the Tennessee River, February 27, 2023
71124.07 - Radiological Environmental Monitoring Program
Environmental Monitoring Equipment and Sampling (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated environmental monitoring equipment and observed collection of environmental samples.
Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program.
GPI Implementation (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees implementation of the Groundwater Protection Initiative (GPI) program to identify incomplete or discontinued program elements.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05)===
- (1) Unit 1 (July 1, 2023, through June 30, 2024)
- (2) Unit 2 (July 1, 2023, through June 30, 2024)
- (3) Unit 3 (July 1, 2023, through June 30, 2024)
MS09: Residual Heat Removal Systems (IP Section 02.08) (3 Samples)
- (1) Unit 1 (July 1, 2023, through June 30, 2024)
- (2) Unit 2 (July 1, 2023, through June 30, 2024)
- (3) Unit 3 (July 1, 2023, through June 30, 2024)
MS10: Cooling Water Support Systems (IP Section 02.09) (3 Samples)
- (1) Unit 1 (July 1, 2023, through June 30, 2024)
- (2) Unit 2 (July 1, 2023, through June 30, 2024)
- (3) Unit 3 (July 1, 2023, through June 30, 2024)
PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) March 18, 2023, through August 2, 2024
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) CR 1742531, Unit 2 high pressure coolant injection steam exhaust pressure rose to 160 psig on September 9, 2024
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Follow up (IP Section 03.01)
- (1) The inspectors evaluated the loss of alternate decay heat removal and licensees response on September 11, 2024.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.
- (1) LER 05000260/2023-002-00 for Browns Ferry Nuclear Plant, Unit 2, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints (ADAMS Accession No. ML23216A176). The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER; therefore, no performance deficiency was identified. The inspectors determined that this LER was related to a Severity Level IV violation that is documented in the inspection results section of this report. This LER is Closed.
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated operator performance during a dual reactor recirculation pump trip following an automatic reactor scram that occurred on April 24,
INSPECTION RESULTS
Licensee-Identified Non-Cited Violation 71124.06 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: On February 27, 2023, the licensee discovered that the demineralized water system, which is normally free from radioactivity, had been cross contaminated with water containing radioactive material from an operating reactor unit due to backflow through a temporary hose connection. Subsequently, it was determined the demineralized water system was leaking at two points outside of the Unit 3 reactor building and into a storm drain leading to the Tennessee River, thereby bypassing the point-of-discharge dilution flow normally used for radioactive liquid effluent releases. The licensee estimated that the system became contaminated up to nine days prior to discovery and anywhere from 500 to 13,000 gallons of contaminated water was released prior to the leaks being stopped. A licensee investigation determined that a similar cross contamination event occurred in 2005, and that corrective actions from that event (including installation of backflow preventers at demineralized water taps) were never fully implemented. Title 10 of the Code of Federal Regulations (CFR) Part 20.1406 states that licensees shall, to the extent practical, conduct operations to minimize the introduction of residual radioactivity into the site. Contrary to this, sometime prior to February 27, 2023, the licensee failed to conduct operations, to the extent practical, to minimize the introduction of radioactivity into the site. No regulatory limits were exceeded, and this event was described as an abnormal release in the 2023 Annual Radioactive Effluent Release Report.
Significance/Severity: Green. The finding was not the result of a substantial failure to implement the effluent monitoring program and doses to the public did not exceed 10 CFR 50, Appendix I criteria or 10 CFR 20.1301 limits.
Corrective Action References: CR 1840376 Failure to Maintain Reactor Coolant System Thermal Limit Parameters within Technical Specification Requirements Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000259/2024003-01 Open/Closed
[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 1 Technical Specification (TS) 5.4.1.a, "Procedures," was identified when the licensee failed to implement written procedures as specified in Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2, Appendix A, Section 5. Specifically, following the Unit 1 scram on April 24, 2024, the licensee failed to implement the direction of abnormal operating procedures 1-AOI-100-1 and 1-AOI-68-1 to either restart the recirculation pumps or initiate a plant cooldown to maintain compliance with TS surveillance requirement 3.4.9.1, reactor coolant system pressure and temperature limits.
Description:
Browns Ferry Nuclear Unit 1 TS surveillance requirement (SR) 3.4.9.1(1)requires, in part, that both reactor coolant system (RCS) pressure and RCS temperature are maintained within the limits specified by curve number 1 of figures 3.4.9-1 and that RCS heatup and cooldown rates are less than or equal to 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
On April 24, 2024, at 10:15 p.m. Central Standard Time, Unit 1 automatically scrammed due to a faulted 1B main bank transformer. Because this transformer powers normal Unit 1 station auxiliary equipment, this transformer failure caused a loss of the Unit 1 condensate and condensate booster pumps. The loss of these pumps led to a low suction trip of all three feedwater pumps. This loss of condensate and feedwater flow caused reactor pressure vessel (RPV) level to lower to the point where the recirculation pumps tripped and both the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) pumps started, as designed, to restore RPV water level. The loss of this transformer also resulted in a loss of non-essential main control room computers used by the operators to monitor various plant parameters. Following the trip of both reactor recirculation pumps, operators entered abnormal operating procedures 1-AOI-100-1, Reactor Scram, 1-AOI-68-1, Recirc Pump Trip/Core Flow Decrease, and 0-AOI-57-1B, Loss of 500KV. Both 1-AOI-68-1 and 1-AOI-100-1 direct that if the recirculation pumps could not be started, then initiate a plant cooldown to ensure the limits of TS figure 3.4.9-1 are not exceeded. A caution statement in both procedures also warned that if the recirculation delta temperature limits are exceeded and a cooldown is not started and continued, the pressure limit for the reactor vessel bottom head will be exceeded.
The loss of the non-essential control room computers contributed to a delay in the operators restart of the recirculation pumps. Normally post-scram, operators use an automatic cooldown feature of the turbine bypass valve control logic where a cooldown rate is set by the operators and the bypass valves modulate to maintain the set rate. Because of complications with the transformer fault, the automatic cooldown feature of the turbine bypass valves was not immediately available. As the operators were focused on restoring the automatic cooldown feature, they did not promptly start a cooldown even though several other means were available to establish a manual cooldown. For example, operators could have used procedure 1-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations, to manually cooldown using the main turbine bypass valves.
With no recirculation pumps operating and relatively cool control rod drive water entering the bottom head region of the reactor vessel, thermal stratification occurred. This allowed the bottom head temperatures to fall below the minimum allowable temperature as per curve 1 of TS figure 3.4.9-1, unbeknownst to the operators.
Approximately five hours after the scram, operators began monitoring reactor heatup and cooldown as required by 1-SR-3.4.9.1(1), "Reactor Heatup and Cooldown Rate Monitoring,"
Revision 14. The operator performing 1-SR-3.4.9.1(1) was also involved in many other duties such as making log entries, printing procedures, answering phones, and briefing auxiliary unit operators. This operator recorded the 1-SR-3.4.9.1(1) data but failed to notice that acceptance criteria steps were not being met. 1-SR-3.4.9.1(1) step 5.5 states, in part, to review the data and ensure readings on attachment 1 comply with curve 1 of figure 3.4.9-1 and to sign-off the acceptance criteria step. Because this operator was unaware that acceptance criteria steps were not met, the nuclear unit senior operator, the shift technical advisor, and the shift manager were also unaware of the failure to comply with curve 1.
The following morning, the day shift crew resolved the issues with the automatic cooldown feature and commenced the cooldown. During the evolution, relatively high temperature feedwater was introduced into the bottom head restoring flow to the stratified region. This resulted in a heatup rate of approximately 185°F/hr. When the 1-SR-3.4.9.1(1) data was reviewed, operators identified the failure to meet both the minimum required temperature of curve 1 and the maximum 100°F RCS heatup rate. As required by TS 3.4.9.A.2, the licensee contracted with a third-party vendor to perform an analysis to determine if the RCS was acceptable for operation. This evaluation determined that RCS structural integrity was not adversely affected, and that the RCS was acceptable for continued operation. The inspectors reviewed the licensee's evaluation and performed an independent evaluation and agreed with the licensee's conclusion.
Corrective Actions: The licensee restored parameters to within limits and determined that the RCS was acceptable for continued operation via an engineering evaluation.
Corrective Action References: CR 1927039
Performance Assessment:
Performance Deficiency: The licensees failure to implement the direction of abnormal operating procedures 1-AOI-100-1 and 1-AOI-68-1 to either restart the recirculation pumps or initiate a plant cooldown to maintain compliance with SR 3.4.9.1 RCS pressure and temperature limits was a performance deficiency. Specifically, not adhering to the caution statement in both procedures caused the RCS pressure and temperature limits of TS SR 3.4.9.1 to be exceeded.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the condition affected the reactor vessel bottom head region integrity and required an engineering evaluation to verify that continued RCS operation was acceptable.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. The inspectors screened the performance deficiency using Exhibit 4, "Barrier Integrity Screening Questions,"
and determined that the issue screened to Green based on answering "No" to all the screening questions.
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the operator tasked with monitoring RCS heatup and cooldown rates failed to perform a thorough review of the task rather than relying on past successes and assumed conditions. Additionally, the responsible nuclear unit senior operator failed to ensure specific contingency actions of both AOIs were discussed and understood with the reactor operators responsible for plant recovery.
Enforcement:
Violation: Browns Ferry Nuclear Unit 1 Technical Specification 5.4.1.a states, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2, Appendix A, Section 5 recommends procedures for abnormal conditions. The licensee established abnormal operating instructions 1-AOI-100-1 and 1-AOI-68-1 to provide instructions on plant recovery following a reactor scram and a recirculation pump trip.
Contrary to the above, on April 24 and 25, 2024, the licensee failed to implement written procedures 1-AOI-100-1 and 1-AOI-68-1 for abnormal conditions following a reactor scram and recirculation pump trip. Specifically, the licensee failed to either restart the recirculation pumps or initiate a plant cooldown following the scram to maintain compliance with RCS pressure and temperature limits of Technical Specification Surveillance Requirement 3.4.9.1(1).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000260/2024003-02 Open/Closed Not Applicable 71153 A self-revealed Severity Level IV non-cited violation of Technical Specifications 3.4.3 and 3.0.4 was identified when the licensee discovered, through as-found test results, that four of thirteen main steam relief valves removed for testing had as-found lift settings outside of the
+/-3 percent setpoint band required for operability.
Description:
Browns Ferry Nuclear (BFN) Unit 2 Technical Specification 3.4.3 requires twelve of the thirteen main steam relief valves (MSRVs) to be operable while in Modes 1, 2, and 3.
On June 5, 2023, the licensee was notified of as-found testing results that four MSRVs from Unit 2 were outside of the +/-3 percent setpoint band required for operability per surveillance requirement 3.4.3.1. It was determined that MSRV BFN-2-PCV-001-0004 failed due to corrosion bonding to the valve seat and simmering. MSRV BFN-2-PCV-001-0019, BFN-2-PCV-001-0031, and BFN-2-PCV-001-0179 failed due to a relaxation of the setpoint spring over time. All MSRVs were considered to be inoperable during the entire operating cycle from April 21, 2021, to February 17, 2023, and longer than permitted by TS 3.4.3. Additionally, TS 3.0.4 requires that when a limiting condition for operation (LCO) is not met, entry into an applicable mode or specified condition is not permitted unless the associated actions permit continued operation. On April 21, 2021, BFN Unit 2 entered a TS 3.4.3 applicable mode when TS LCO 3.4.3 Required Actions were not met.
The affected valves remained capable of maintaining reactor pressure below the ASME Code limit of 1375 psig. All thirteen of the MSRV pilot valves were replaced during the Unit 2 spring 2021 refueling outage. The previous corrective action from LER 05000260/2021-002-00 to apply a platinum coating to the pilot using the plasma enhanced magnetron sputtering deposition method (PEMS), which improves the quality and adhesion of the coating, has been utilized. A flaking issue has been noted with the platinum coated pilot disc. The Boiling Water Reactor Owners' Group is currently working toward a solution to improve the quality and adhesion of the platinum coating on the discs.
Corrective Actions: The licensee replaced all thirteen MSRV pilot valves during the spring 2023 (U2R22) refueling outage. As-left testing verified that these refurbished pilot valves were within +/- 1 percent of their setpoints. The installed valves have implemented corrective actions from past occurrences of corrosion bonding that include preparing the pilot discs in accordance with the revised procedure and vendor recommendations. The currently installed refurbished valves had platinum coatings applied utilizing the PEMS deposition method, and as-left values were verified to be within +/- 1 percent of their setpoints. Additional corrective actions may be developed based on feedback from the Boiling Water Reactor Owners Group related to corrosion bonding of this specific model safety relief valve.
Corrective Action References: CRs 962223, 1286467, 1410577, 1521190, 1658693, 1699286, 1775232, 1822254, and 1860559
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Enforcement:
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.
Severity: This violation is characterized as a Severity Level (SL) IV violation based on its similarity to SL IV example 6.1.d.1 in the Enforcement Policy. The inspectors also reviewed NRC Enforcement Policy, Section 2.2.1, "Factors Affecting Assessment of Violations," which states, in part, that in determining the appropriate enforcement response to a violation, the NRC considers, whenever possible, risk information in assessing the safety or security significance of violations and assigning severity levels. The inspectors determined the issue to be of very low safety significance because the valves remained capable of performing their required safety function.
Violation: Browns Ferry Nuclear Unit 2 Technical Specification 3.4.3, "Safety/Relief Valves (S/RVs)," Condition A, requires that with one or more required S/RVs inoperable, the unit be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Contrary to the above, four required Unit 2 S/RVs were inoperable from April 21, 2021, to February 17, 2023, and the unit did not enter Mode 3 and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively.
Browns Ferry Nuclear Unit 2 Technical Specification LCO 3.0.4 requires, in part, that when an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time.
Contrary to the above, on April 21, 2021, Unit 2 entered a Technical Specification 3.4.3 applicable mode when Technical Specification LCO 3.4.3 required actions were not met.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On August 1, 2024, the inspectors presented the public radiation safety baseline inspection results to Daniel Komm, Site Vice President, and other members of the licensee staff.
- On September 12, 2024, the inspectors presented the in-service inspection results to Daniel Komm, Site Vice President, and other members of the licensee staff.
- On October 17, 2024, the inspectors presented the integrated inspection results to Daniel Komm, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or Date
Drawings
3-47E811-1
Flow Diagram Residual Heat Removal System
Revision 80
Miscellaneous
BFN-VTD-T147-
0500
Turbine Governor Control System HPCI CS and CCS
Revision 0
Miscellaneous
BFN-VTD-T147-
0510
Governing Oil Relay and Linkage System
Revision 0
Procedures
3-OI-74/ATT-1
Residual Heat Removal System Valve Lineup Checklist
April 28, 2008
Procedures
3-OI-74/ATT-2
Residual Heat Removal System Panel Lineup Checklist
Revision 90
Procedures
3-OI-74/ATT-3
Residual Heat Removal System Electrical Lineup
Checklist
Revision 86
Procedures
EPI-0-077-
SWZ002
Functional Check of the Reactor Building Flood Level
and Equipment Access Lock Water Seal Level Switches
Revision 9
Corrective Action
Documents
21501,
22121,
1736146,
1771965,
1794387,
1800115,
1813881,
1579199,
22886,
1772413,
1850940,
1941210,
1941338,
1941181,
1941059,
1874485,
1869439,
1939735,
1871312,
1788114,
1787594,
Inspection
Procedure
Type
Designation
Description or Title
Revision or Date
1945943
Miscellaneous
Function 071-B:
(a)(1) Evaluation
Miscellaneous
Purchase order
6909657, Procure
six rupture discs
July 13, 2021
Miscellaneous
R40 230803 463
Maintenance Rule
Expert Panel
Meeting Minutes
Miscellaneous
R40 240229 549
Maintenance Rule 14th Periodic Report
Revision 0
Procedures
0-TI-384
CAD Tank Boil-Off Determination
Revision 10
Work Orders
21305712,
21305714,
23803867,
23803644,
2130572,
21305714
Corrective Action
Documents
CR 1942352,
1942357
Corrective Action
Documents
CR 1942327,
1955855,
1742531,
1952331,
1957254,
1954800,
1954970,
1954797
Drawings
1-730E321
Sh. 3, "Elementary Drawing Reactor Manual Control
System"
Revision 8
Miscellaneous
Selected
Dataware plots
for Unit 1 & 2
September 9,
24
Miscellaneous
Selected main
August 30, 2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or Date
control room
operator logs
Miscellaneous
Selected main
control room
operator logs
September 9,
24
Engineering
Changes
DCN 71458
Add support to lift residual heat removal service water
cables off floor of handholes 26 and 15
Revision A
Corrective Action
Documents
CR 1941210,
1941338,
1941181, 956790,
1958139
Drawings
Drawing 0-
Wiring Diagram Door Interlock & Alarm System
Schematic Diagram
Revision 22
Work Orders
23803644,
23794697,
23783014,
23783005,
23848434,
23544185,
23989262,
23236511,
24792337,
24808664,
23643965,
245543462,
24804993,
23544221,
23644012,
24154786
Calculations
Liquid discharge dose calculation, abnormal release, 1st
quarter 2023
Corrective Action
Documents
CR# 1944567
Inspection
Procedure
Type
Designation
Description or Title
Revision or Date
Corrective Action
Documents
PER# 79075
Corrective Action
Documents
CR# 1661007
ODCM Meteorological Data Review
Corrective Action
Documents
CR# 1661936
Remove REMP air sampling location RM-1 from ODCM
Corrective Action
Documents
CR# 1702786
REMP Water Station 293.5 found with no sample volume
during checks
Corrective Action
Documents
CR# 1708729
REMP air station LM-2 found not operating
Corrective Action
Documents
CR# 1726843
Document informal BM of Groundwater sample
Corrective Action
Documents
CR# 1744436
Corrective Action
Documents
CR# 1790567
REMP Week 27 water samples lost in transit
Corrective Action
Documents
CR# 1838768
Gamma Activity in Weekly ADHR sample
Corrective Action
Documents
CR# 1840334
Power unavailable for REMP LM-4 air sample station
Corrective Action
Documents
CR# 1840376
Unmonitored leak of greater than 100 gallons to the
environment
Corrective Action
Documents
CR# 1853201
10CFR50.75(g) Documentation for Contaminated Demin
Water System
Corrective Action
Documents
CR# 1884208
REMP 3Q Dosimeter SSW-2 missing
Corrective Action
Documents
CR# 1903368
REMP PM-3 air sample unable to be collected
Corrective Action
Documents
CR# 1907957
Failed As-Found Calibration of 10m, 46m RTDs at BFN
Miscellaneous
50.75(g) Demin
Water Leak
2/27/2023
50.75(g) Demin Water Leak 02/27/2023 Report
May 31, 2023
Corrective Action
CR 1927039
Inspection
Procedure
Type
Designation
Description or Title
Revision or Date
Documents