IR 05000259/2023003

Integrated Inspection Report 05000259/2023003, 05000260/2023003 and 05000296/2023003

ML23312A351
Person / Time
Site:	Browns Ferry
Issue date:	11/13/2023
From:	Louis Mckown Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
IR 2023003
Download: ML23312A351 (1)

Text

November 13, 2023

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2023003, 05000260/2023003 AND 05000296/2023003

Dear Jim Barstow:

On September 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On October 5, 2023, the NRC inspectors discussed the results of this inspection with Mr. Darrell Lock, Assistant Plant Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.

Five findings of very low safety significance (Green) are documented in this report. Five of these findings involved violations 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.

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. 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, Signed by McKown, Louis on 11/13/23 Lou J. McKown, II, Chief Reactor Projects Branch #5 Division of Reactor Projects Docket Nos. 05000259, 05000260 and 05000296 License Nos. DPR-33, 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: 05000259/2023003, 05000260/2023003 and 05000296/2023003 Enterprise Identifier: I-2023-003-0013 Licensee: Tennessee Valley Authority Facility: Browns Ferry Nuclear Plant Location: Athens, Alabama Inspection Dates: July 01, 2023, to September 30, 2023 Inspectors: S. Billups, Project Engineer N. Karlovich, Resident Inspector A. Nielsen, Senior Health Physicist K. Pfeil, Resident Inspector T. Steadham, Senior Resident Inspector Approved By: Lou J. McKown, II, Chief Reactor Projects Branch #5 Division of Reactor Projects Enclosure

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.

List of Findings and Violations

Failure to Identify Non-Functional Fire Door in the Corrective Action Program Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71111.05 Systems NCV 05000296/2023003-01 Complacency Open/Closed The inspectors identified a Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Renewed Facility Operating License, DPR-68, Condition 2.C.(7) when the licensee failed to identify nonfunctional fire doors in the corrective action program. As a result, the licensee failed to take appropriate corrective actions, such as implementing the required compensatory measures, until identified by the NRC.

Failure to Comply with Radiation Work Permit Requirements for Entry into a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV Teamwork 05000259,05000260,05000296/202300 3-02 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Units 1, 2, and 3 Technical Specification 5.4.1.a was identified when a worker failed to comply with radiation work permit requirements for entry into a high radiation area. Specifically, the worker entered the Unit 2 fuel pool cooling cage area, a posted high radiation area with dose rates exceeding 100 mrem/hr at 30 cm, but less than 1,000 mrem/hr at 30cm, without receiving a required briefing from radiation protection.

Failure to Maintain Standby Liquid Control Chemistry Parameters within Technical Specification Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71153 Systems NCV 05000296/2023003-03 Conservative Open/Closed Bias A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Technical Specification 5.4.1.a was identified when the licensee failed to maintain the standby liquid control system conditions within the parameters required by Technical Specification Surveillance Requirement 3.1.7.6. Specifically, the combination of standby liquid control tank sodium pentaborate concentration, Boron-10 enrichment, and standby liquid control pump flowrate was insufficient to support operability of both Unit 3 standby liquid control pumps during as-found surveillance testing on October 18, 2022.

Pressure Boundary Leak on Reactor Recirculation Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000260/2023003-04 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 2 Technical Specification 3.4.4 and 10 CFR 50, Appendix B, Criterion III, "Design Control," was identified when the licensee failed to accurately model the supports needed for the drain pipe assembly for the Unit 2 recirculation pump A discharge valve, 2-FCV-068-0003. As a result, a pressure boundary leak developed when flow induced vibrations caused cyclic fatigue failure at the location.

Pressure Boundary Leak on Shutdown Cooling Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000296/2023003-05 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Technical Specification 3.4.4 and 10 CFR 50, Appendix B, Criterion III, "Design Control," was identified when the licensee failed to accurately model the supports needed for small bore piping associated with the Unit 3 loop I residual heat removal shutdown cooling line. As a result, a pressure boundary leak developed on a vent line for this shutdown cooling line when flow induced vibrations caused cyclic fatigue failure at the location.

Browns Ferry Nuclear Plant, Unit 1, Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259/2023003-06 Open/Closed A self-revealed Severity Level IV non-cited violation of Browns Ferry Nuclear Unit 1 Technical Specifications 3.4.3 and LCO 3.0.4 was identified when the licensee discovered, through as found test results, that two of thirteen main steam relief valves that were removed for testing had as found lift settings outside of the +/-3 percent setpoint band required for their operability.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000259/2023-001-01 LER 2023-001-01 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 1, High Pressure Coolant Injection System

Inoperable Due to a Torn Valve Diaphragm LER 05000259/2023-001-00 LER 2023-001-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 1, High Pressure Coolant Injection System Inoperable Due to a Torn Valve Diaphragm LER 05000296/2022-002-00 LER 2022-002-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 3, Both Standby Liquid Control Subsystems Inoperable Due to an Insufficient Boron Injection Rate LER 05000296/2022-003-00 LER 2022-003-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 3, Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line LER 05000259/2022-003-00 LER 2022-003-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 1, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints LER 05000260/2023-001-00 LER 2023-001-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 2, Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line LER 05000296/2022-003-01 LER 2022-003-01 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 3, Pressure Boundary Leak on Residual Heat

Removal System Low Pressure Coolant Injection Test Line due to Fatigue Failure LER 05000260/2023-001-01 LER 2023-001-01 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 2, Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line Due to Fatigue Failure

PLANT STATUS

Unit 1 began the inspection period at full (100 percent) rated thermal power (RTP). On September 2, 2023, operators lowered reactor power to 67 percent RTP for a control rod sequence exchange. On September 3, 2023, the unit was returned to 100 percent RTP where it operated at or near for the remainder of the inspection period.

Unit 2 began the inspection period at full RTP. On September 1, 2023, operators lowered reactor power to 67 percent RTP for a control rod sequence exchange. On September 2, 2023, the unit was returned 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 July 21, 2023, operators lowered reactor power to 62 percent RTP for a control rod sequence exchange and main turbine condenser waterbox cleaning. On July 23, 2023, the unit was returned to 100 percent RTP. On September 8, 2023, operators lowered reactor power to 68 percent RTP for a control rod sequence exchange. On September 10, 2023, 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 on-site 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.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (4 Samples)

(1) Unit 3 emergency diesel generator 3B starting air, fuel oil, electrical, and panel lineup during 3A emergency diesel generator planned maintenance on July 25, 2023.

(2) Unit 1 residual heat removal loop 2 valve, panel, and electrical lineup during planned loop 1 maintenance on August 3, 2023

(3) Unit 1 core spray loop 1 while loop 2 unavailable for planned maintenance on August 11, 2023

(4) Unit 1 residual heat removal loop 1 valve, panel, and electrical lineup during planned core spray loop 2 maintenance on August 16, 2023

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configuration during a complete walkdown of the Unit 1 250 VDC distribution system on September 21, 2023

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (11 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) Fire area 01-01, Unit 1 reactor building elevation 565', on July 12, 2023

(2) Fire area YARD, areas outside Units 1, 2, and 3 emergency diesel generators, on July 14, 2023

(3) Fire area 21, Unit 3 emergency diesel generator building, elevation 565', on July 25, 2023

(4) Fire area 27, Units 1 and 2 A/B control bay chiller rooms on August 2, 2023

(5) Fire area 2-2, Unit 2 residual heat removal loop 2 pump room, elevation 541' on August 24, 2023

(6) Fire area REFUEL, Unit 1, 2, and 3 refueling floor, elevation 664' on August 30, 2023

(7) Fire area 16, Unit 3 main control room on September 08, 2023

(8) Fire area YARD, channel diesel fire pump building on September 08, 2023

(9) Fire area 16, Unit 1 1C control bay hallway, 593' elevation on September 13, 2023

(10) Fire Area 17, Unit 1 battery and battery board room, control bay elevation 593' on September 18, 2023

(11) Fire area 16, Unit 2 cable spreading room B, control bay elevation 606' on September 25, 2023

Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill located in the Unit 3 ventilation room, control bay elevation 606' on August 22, 2023

71111.06 - Flood Protection Measures

Flooding Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated external flooding mitigation protections for the 565' elevation of the Units 1, 2, and 3 emergency diesel generator buildings on September 07, 2023

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Unit 3 3A emergency diesel generator jacket water cooling heat exchangers 3A and 3B inspection and replacement under work order (WO) 123068155 on July 27, 2023

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) Operator performance during Unit 2 rod pattern adjustment on September 06, 2023

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) Licensed operator requalification examination in the Unit 3 simulator on September 05, 2023

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 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) Maintenance rule monitoring program periodic assessment report review on September 14, 2023

(2) Unit Common control bay chillers and control bay system maintenance rule monitoring on September 28, 2023

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) Procurement and receipt inspection documents for replacement Unit 3 emergency diesel generator 3A jacket water heat exchangers on July 27, 2023.

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (7 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) Protected equipment lineup while Unit 3 3A emergency diesel generator unavailable on July 28, 2023

(2) Protected equipment lineup due to Units 1/2 A control bay chiller unavailable on July 31, 2023

(3) Protected equipment lineup due to Unit 1 loop 2 core spray unavailable on August 11, 2023

(4) Protected equipment lineup while Unit 2 loop 1 residual heat removal unavailable on August 24, 2023

(5) Protected equipment lineup while Unit common standby gas treatment train C unavailable on September 13, 2023

(6) Protected equipment lineup while Unit 3 3B emergency diesel generator unavailable on September 25, 2023

(7) Maintenance risk and emergent work control during motor replacement of the Unit 1 1A reactor protection system motor generator set on September 29, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Condition report (CR) 1872763, Unit 3 emergency diesel generator building flooding on August 7, 2023

(2) CR 1873186, Radiation recorder 0-RR-90-147 channel found out of tolerance on August 10, 2023

(3) CR 1875761, Unit Common control bay chiller operability determination due to the roof structure installed by engineering change DCN-69983 on September 06, 2023

(4) CR 1878492, Down stream piping of high pressure coolant injection steam line outboard isolation valve would not depressurize on September 07, 2023

(5) CR 1879992, Failure to perform periodic maintenance on Unit 3 high pressure coolant injection inboard torus suction valve, 3-MVOP-073-0026 on September 13, 2023

(6) CR 1880541, Unit 2 2A residual heat removal pump oil leak on September 18, 2023

(7) CR 1880309, Unit common standby gas treatment damper, BFN-0-FCO-065-0051, failed smoke test on September 19, 2023

(8) CR 1880021, Unit 1/2 emergency diesel generator C air dryer needs replacing on September 25, 2023

(9) CR 1883072, possible bent seismic support for Unit 3 3EB 250 VDC battery cell on September 29, 2023

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) BFN-2-2023-074-001, Unit 2 cut and cap test line for residual heat removal test valves on August 16, 2023

(2) Revision 2 to calculation MDQ007320100031, high pressure coolant injection pump and system hydraulic analysis, as a result of valve modifications on September 28, 2023

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)

(1) WO 120151834, replace Unit common standby gas treatment B main control room fan switch, BFN-0-HS-065-0040A/1, on August 3, 2023

(2) WO 121469373, replace Unit common 480V diesel aux board B standby gas treatment B relay, BFN-0-RLY-065-0034A, on August 3, 2023

(3) WO 123906037, Unit 1 residual heat removal pump A shutdown cooling suction valve would not operate during testing, BFN-1-MVOP-074-0002, on August 10, 2023

(4) WO 118621383, replace Unit 1 1A control rod drive pump motor, BFN-1-MTR-085-0001, on August 18, 2023

(5) WO 121597617, perform diagnostic testing on Unit 3 high pressure coolant injection outboard torus suction valve, 3-MVOP-073-0027 on September 12, 2023

(6) WO 123169346, lubricate and inspect Unit 3 high pressure coolant injection suppression pool inboard suction valve, BFN-3-MVOP-073-0026 on September 19, 2023

(7) WO 119334041, replace the stem and wedge pin on Unit 1 core spray minimum flow isolation valve, BFN-1-FCV-075-0009 on September 22, 2023

(8) WO 123808713, replace valve seat in Unit 1 control rod scram outlet valve, BFN-1-FCV-085-39B, on September 26, 2023

(9) WO 123808788, replace valve seat in Unit 1 control rod scram inlet valve, BFN-1-

FCV-085-39A, on September 26, 2023 Surveillance Testing (IP Section 03.01) (4 Samples)

(1) WO 122954769, Unit 3 Diesel Generator 3A Monthly Operability Test, 3-SR-3.8.1.1(3A), on July 7, 2023.

(2) WO 121471563, system leakage test of the Unit 3 reactor pressure vessel and associated piping on September 01, 2023.

(3) WO 123153907, Unit 2 high pressure coolant injection developed head and flow rate test at rated reactor pressure on September 15, 2023

(4) WO 123030205, quarterly check for the 250 VDC main bank number 1 battery on

September 21, 2023 Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) WO 123071652, Unit 1 standby liquid control pump surveillance test, 1-SI-4.4.A.1, on August 9, 2023

Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)

(1) WO 123197462, 4160V FLEX generator annual maintenance on September 19, 2023

71114.06 - Drill Evaluation

Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)

The inspectors evaluated:

(1) The inspectors observed and evaluated an emergency preparedness drill on August 2, 2023. Events included an anticipated transient without scram, loss of emergency core cooling pumps, unisolable reactor coolant system leak, loss of containment integrity, and loss of reactor pressure vessel level indication.

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, 2022, through June 30, 2023, on September 14, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023, on September 14, 2023)

(3) Unit 3 (July 1, 2022, through June 30, 2023, on September 14, 2023)

MS09: Residual Heat Removal Systems (IP Section 02.08) (3 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023, on September 14, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023, on September 14, 2023)

(3) Unit 3 (July 1, 2022, through June 30, 2023, on September 14, 2023)

MS10: Cooling Water Support Systems (IP Section 02.09) (3 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023, on September 14, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023, on September 14, 2023)

(3) Unit 3 (July 1, 2022, through June 30, 2023, on September 14, 2023)

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 1820718, pressure boundary leak on Code Class I residual heat removal piping on September 15, 2023

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000259/2022-003-00 for Browns Ferry Nuclear Plant, Unit 1, Main Steam Relief Valve (MSRV) Setpoint Failure (ADAMS Accession No. ML23037A992). 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 inspection conclusions associated with this LER are documented in the Inspection Results section of this report. This LER is Closed.

(2) LER 05000259/2023-001-00 for Browns Ferry Nuclear Plant, Unit 1, High Pressure Coolant Injection System Inoperable Due to a Torn Valve Diaphragm (ADAMS Accession No. ML23086C092). 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 did not identify a violation of NRC requirements. This LER is Closed.

(3) LER 05000259/2023-001-01 for Browns Ferry Nuclear Plant, Unit 1, High Pressure Coolant Injection System Inoperable Due to a Torn Valve Diaphragm (ADAMS Accession No. ML23164A134). This LER was a revision to an LER which was previously inspected. No findings or violations were identified as a result of the review of this revision. This LER is Closed.

(4) LER 05000260/2023-001-01 for Browns Ferry Nuclear Plant, Unit 2, Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line (ADAMS Accession No. ML23261C399). This LER was a revision to an LER which was previously inspected. No additional findings or violations were identified as a result of the review of this revision. This LER is Closed.

(5) LER 05000260/2023-001-00 for Browns Ferry Nuclear Plant, Unit 2, Pressure Boundary Leak on Recirculation Pump Discharge Isolation Valve Drain Line (ADAMS Accession No. ML23109A362). The inspection conclusions associated with this LER are documented in the Inspection Results Section of this report. This LER is Closed.

(6) LER 05000296/2022-002-00 for Browns Ferry Nuclear Plant, Unit 3, Both Standby Liquid Control Subsystems Inoperable Due to an Insufficient Boron Injection Rate, (ADAMS Accession No. ML22353A631). The inspection conclusions associated with this LER are documented under the Inspection Results Section of this report. This LER is Closed.

(7) LER 05000296/2022-003-00 for Browns Ferry Nuclear Plant, Unit 3, Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line (ADAMS Accession No. ML23031A357). The inspection conclusions associated with this LER are documented under the Inspection Results Section of this report. This LER is Closed.

(8) LER 05000296/2022-003-01 for Browns Ferry Nuclear Plant, Unit 3, Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line (ADAMS Accession No. ML23212B015). This LER was a revision to an LER which was previously inspected. No additional findings or violations were identified as a result of the review of this revision. This LER is Closed.

INSPECTION RESULTS

Failure to Identify Non-Functional Fire Door in the Corrective Action Program Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71111.05 Systems NCV 05000296/2023003-01 Complacency Open/Closed The inspectors identified a Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Renewed Facility Operating License, DPR-68, Condition 2.C.(7) when the licensee failed to identify nonfunctional fire doors in the corrective action program. As a result, the licensee failed to take appropriate corrective actions, such as implementing the required compensatory measures, until identified by the NRC.

Description:

On July 24, 2023, the inspectors identified two fire doors that were not meeting one of their intended design functions of providing a sealing barrier in the event of a carbon dioxide (CO2) discharge. The inspectors found doors 3-DOOR-260-0808 (door 808) and 3-DOOR-260-0809 (door 809) closed but not latched. While inspecting both doors, the latching mechanisms for both doors were sticking and would not consistently latch when either door was closed. These doors were relied upon as barriers to secure CO2 in the C and D emergency diesel generator (EDG) rooms, respectively. The inspectors immediately informed the licensee of their concern with the doors.

The following day, the inspectors noted that a condition report had not yet been written for either door. The inspectors performed another walk down and found that while door 808 was latched, door 809 was closed but not latched. The latching mechanism for door 809 was again sticking and would not consistently latch when the door was closed. After notifying the licensee, the licensee determined that both latching mechanisms were degraded beyond the scope of tool pouch maintenance, and both needed more extensive repairs to restore full functionality. As required by fire protection LCO 9.3.11.D.2, the licensee implemented a one-hour fire watch.

The licensees Fire Protection Requirements Manual, Revision 21, Section 2.4, required that for both doors to remain functional, they must also latch because both doors opened out of the room the CO2 system was designed to protect. The requirement to latch was also included in the various implementing procedures such as 0-SI-4.11.G.2.b, Fire Door Inspection, Revision 26. Based on these requirements, the licensee determined that the doors not meeting the design requirement would impair the CO2 system design function to discharge and maintain the required CO2 concentration to extinguish a fire.

The inspectors reviewed CR history for door 809 and identified CR 1828397, written on January 10, 2023, documenting door 3-DOOR-260-0809 was not latching properly and that the latching mechanism needed to be replaced. The licensee found that the latching mechanism for door 809 was sticking and would not consistently latch in the same manner as the inspectors found on both July 24 and 25, 2023. Corrective actions for CR 1828397 consisted of tool pouch maintenance to repair the door latching mechanism. Because the licensee believed the door was repaired, no compensatory measures were implemented. After speaking with licensee personnel responsible for repairing these doors, the inspectors learned that an unknown number of additional repairs to the latching mechanism were frequently performed under tool pouch maintenance - none of which were documented in the corrective action program The inspectors questioned why a CR was not written until July 25, 2023, on these two doors after the inspectors brought the issue up a second time. In discussion with site staff, the inspectors learned that although these doors were routinely found to be not properly latching, the standard practice was to not enter the issue into the corrective action program if the door could be fixed via tool pouch maintenance. Consequently, pertinent trends were not identified to alert the licensee that the tool pouch maintenance being frequently performed on the latching mechanism was not fixing the underlying problem.

Licensee procedure NPG-SPP-18.4.5, Fire Protection Quality Assurance, Revision 3, Section 3.2.9 required that malfunctioning fire protection features shall be handled in accordance with NPG-SPP-22.300, Corrective Action Program. Revision 24 of NPG-SPP-22.300, Section 2.0.C and 3.2.C required the initiation of CRs for adverse conditions in accordance with NPG-SPP-01.16, Condition Report Initiation. Revision 6 of NPG-SPP-01.16, Step 3.2.3 provided an example of malfunctioning fire doors as an example of an issue which required the initiation of a CR.

The inspectors concluded that both doors were likely non-functional since January 10, 2023, and that multiple unsuccessful attempts were made to fix the doors using tool pouch maintenance without being documented in the corrective action program.

Corrective Actions: The licensee implemented an hourly fire watch.

Corrective Action References: CR 1870321

Performance Assessment:

Performance Deficiency: The failure to identify nonfunctional CO2 doors in the corrective action program and to implement compensatory measures, as required by the fire protection program, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, unlatched fire doors could affect the availability, reliability, and capability of the CO2 system in response to a fire in the diesel rooms.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. The finding was assessed a high degradation rating because it represented a hole in the wall greater than specified in the low degradation section. Because the finding represented a degraded or non-functional fixed suppression system that adversely affect the ability of the system to protect equipment important to safe shutdown, the inspectors screened the finding based on the licensees fire PRA results. Conservatively assuming that both doors remain unlatched between January 10, 2023, and July 25, 2023, it was assumed that both CO2 barrier doors and the CO2 systems failed. To evaluate the risk change due to doors unlatching condition, multicompartment fire scenarios for EDG room C (PAU 21-C) to PAU 21-E (through door 808) and EDG room D (PAU 21-D) to PAU 21-E (through door 809) were modeled and quantified. For both PAUs 21-C and 21-D, the EDG oil fire is the only fire event that could produce a hot gas layer resulting in multicompartment analysis fire scenarios. Based on these assumptions, the maximum change in core damage frequency was 1.27E-9 and the maximum change in LERF was 3.24E-11. Based on these results, the performance deficiency screened as Green.

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, individuals performing tool pouch maintenance on the doors expected the condition to be corrected but did not plan for the possibility that the deficient condition with the latching mechanisms required more invasive corrective actions.

Enforcement:

Violation: Browns Ferry Nuclear (BFN) Unit 3 Renewed Facility Operating License, DPR-68, Condition 2.C.(7) requires, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility. BFN Fire Protection Report Part V, Program Documentation, Configuration Control, and Quality, Section 5.3.1 referenced procedure NPG-SPP-18.4.5. NPG-SPP-18.4.5, Revision 3, Section 3.2.9 stated, in part, that Nonconforming Items and Corrective Actions Nonconforming, inoperative, or malfunctioning features including adverse conditions in program administration shall be handled in accordance with NPG-SPP-22.300, Corrective Action Program. NPG-SPP-22.300 states, in part, that individuals identify a condition by initiation of a condition report in accordance with NPG-SPP-01.16. NPG-SPP-01.16, Revision 6, Section 3.2.3 requires the initiation of a condition report for malfunctioning fire doors.

Contrary to the above, from January 10, 2023, until July 25, 2023, the licensee failed initiate a condition report for malfunctioning fire doors.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Comply with Radiation Work Permit Requirements for Entry into a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV Teamwork 05000259,05000260,05000296/20230 03-02 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Units 1, 2, and 3 Technical Specification 5.4.1.a was identified when a worker failed to comply with radiation work permit requirements for entry into a high radiation area. Specifically, the worker entered the Unit 2 fuel pool cooling cage area, a posted high radiation area with dose rates exceeding 100 mrem/hr at 30 cm, but less than 1,000 mrem/hr at 30cm, without receiving a required briefing from radiation protection.

Description:

On July 27, 2023, an auxiliary operator performed rounds in various plant areas. The worker was logged in on radiation work permit 23220552 with self-reading dosimeter alarm set points of 60 mrem and 150 mrem/hr, and had been briefed by radiation protection (RP) personnel on radiological conditions and RP controls relevant to the planned work areas. At some point during the rounds the worker received a call to perform additional duties in the Unit 2 fuel pool cooling cage, which had not been part of the original pre-job briefing. The Unit 2 fuel pool cooling cage area was a posted high radiation area with dose rates exceeding 100 mrem/hr at 30 cm, but less than 1,000 mrem/hr at 30cm. Without contacting RP personnel to receive an additional briefing, the worker entered the room and began work activities. The worker subsequently received an unanticipated self-reading dosimeter dose rate alarm of 162 mrem/hr, at which point the worker left the area and reported to RP. The inspectors noted that radiation work permit 23220552 states that a briefing is required prior to every entry into a high radiation area.

Corrective Actions: The licensee took immediate corrective actions including entering the event into their corrective action program and excluding the individual from the radiologically controlled area.

Corrective Action References: CR 1870990

Performance Assessment:

Performance Deficiency: The failure to receive a briefing prior to entry into a high radiation area, as required by radiation work permit 23220552, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, workers who enter high radiation areas without authorization and encounter dose rates exceeding 100 mrem/hr could receive unplanned exposure.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The finding was determined to be of very low safety significance (Green) because it was not related to As Low As Reasonably Achievable planning, did not result in an overexposure, there was no substantial potential for overexposure, and the ability to assess dose was not compromised.

Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, Operations and RP personnel did not clearly coordinate work activities prior to performing work in the Unit 2 fuel pool cooling room area.

Enforcement:

Violation: BFN Units 1, 2, and 3 Technical Specification (TS) 5.4.1.a requires that the procedures recommended by Regulatory Guide 1.33, including procedures for a radiation work permit system, be implemented. Procedure NPG-SPP-05.1, "Radiological Controls" states that each worker must comply with radiation work permit requirements. Radiation work permit 23220552 states that a briefing is required prior to every entry into a high radiation area.

Contrary to the above, on July 27, 2023, a worker entered a high radiation area inside the Unit 2 fuel pool cooling cage without receiving a briefing prior to entry.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Maintain Standby Liquid Control Chemistry Parameters within Technical Specification Requirements Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71153 Systems NCV 05000296/2023003-03 Conservative Open/Closed Bias A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Technical Specification 5.4.1.a was identified when the licensee failed to maintain the standby liquid control system conditions within the parameters required by Technical Specification Surveillance Requirement 3.1.7.6. Specifically, the combination of standby liquid control tank sodium pentaborate concentration, Boron-10 enrichment, and standby liquid control pump flowrate was insufficient to support operability of both Unit 3 standby liquid control pumps during as-found surveillance testing on October 18, 2022.

Description:

The Unit 3 TS requires the verification of various parameters to ensure standby liquid control (SLC) operability such as both the concentration and Boron-10 enrichment of sodium pentaborate (SPB) in the SLC storage tank as well as SLC pump flowrates. For example, surveillance requirement (SR) 3.1.7.4 required verification of the SPB concentration, SR 3.1.7.7 required verification of the SLC pump flow rates, and SR 3.1.7.10 required verification that the Boron-10 enrichment met the limits of a calculation contained in SR 3.1.7.6. Meeting the SR 3.1.7.6 calculation was required to ensure that the SLC system had the capacity to bring the reactor subcritical prior to suppression pool temperature exceeding its heat capacity temperature limit during an anticipated transient without scram with the main steam isolation valves closed.

SR 3.1.7.6 required that the product of the SPB concentration (C), SLC pump flow rate (Q),and Boron-10 enrichment (E), divided by 40,890 be equal to or greater than 1. The licensee developed procedure 3-SR-3.1.7.3 to verify the SPB concentration, 3-SR-3.1.7.7 to verify SLC pump flowrates, and 3-SR-3.1.7.6 to verify adequate enrichment by performing the above calculation.

Because the acceptance criteria contained in 3-SR-3.1.7.3 allowed C to be as low as 8.0 and 3-SR-3.1.7.7 allowed Q to be as low as 48.1 gpm and 49.0 gpm for the 3A and 3B SLC pumps, respectively, the possibility existed where it would be impossible to meet the requirements of SR 3.1.7.10 if both C and Q were at their minimum allowable values because it would require E to exceed 100%. The licensee identified the potential for this occurrence on August 16, 2018, in CR 1439832 and revised procedures 1/2/3-SR-3.1.7.3 and 1/2/3-SI-3.1.7.6; however, these revisions failed to fully consider the interrelationship of the various acceptance criteria to ensure continued SLC operability. As a result, the allowable minimum SPB concentration remained at 8.0 even though this concentration would be inadequate to support operability of the SLC pumps if they were operating at their minimum allowable flowrate.

On July 26, 2022, the licensee performed 3-SI-3.1.7.6 and determined that C was 8.03, Q for the 3A SLC pump was 52.5, and E was 97.64 resulting in a calculated value of 1.007 from 3-SR-3.1.7.6. The licensee identified the reduced margin and added more SPB to the SLC storage tank, increasing the SPB concentration. The as-left SR 3.1.7.6 calculated value was 1.011 on July 27, 2022.

On October 18, 2022, the licensee performed 3-SI-4.4.A.1 and determined that both Unit 3 SLC pumps were operating at 50.6 gpm which met the acceptance criteria for operability. The pumps were restored to operable status following the test. Later that day, the licensee performed procedure 3-SI-3.1.7.6 and determined that C was 8.22 and E was 97.61 yielding a SR 3.1.7.6 calculated value of 0.993. Because both pumps failed SR 3.1.7.6, operators declared both pumps inoperable. The licensee added additional SPB to the SLC storage tank thereby increasing the SPB concentration to a value that satisfied the requirements of SR 3.1.7.6 and restored both pumps to operable within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> required completion time of TS 3.1.7.B.1. On December 19, 2022, the licensee submitted licensee event report 50-296/2022-002-00 to document the identified condition.

Based on the licensees ability to raise SPB concentration after these two occurrences to meet SR 3.1.7.6, the inspectors determined that the licensee could both predict and monitor compliance with SR 3.1.7.6 even with anticipated fluctuations in both measured SLC pump flowrate and Boron-10 enrichment. Because the SLC pump flowrates measured on October 18, 2022, were within the normal expected range of flowrates and with the operating history of low margin to SR 3.1.7.6, the inspectors concluded that the failure to maintain compliance with SR 3.1.7.6 was within the licensees ability to foresee and prevent.

Corrective Actions: The licensee added sufficient SPB to the Unit 3 SLC tank which allowed the successful completion of SR 3.1.7.6 and restored both SLC subsystems to operable. The licensee also revised 1/2/3-SI-4.4.A.1 to modify the acceptance criteria to include considerations related to the SR 3.1.7.6 calculation.

Corrective Action References: CRs 1439832, 1810303, and 1811376

Performance Assessment:

Performance Deficiency: The failure to maintain the SLC system conditions within the parameters required by TS SR 3.1.7.6 was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the condition affected the capability of the Unit 3 SLC to perform its intended function.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the performance deficiency using Exhibit 2 of Appendix A and determined that the issue screened green based on the outage time being less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> required by TS 3.1.7.B.1.

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee did not take action to maintain sufficient margin for SR 3.1.7.6. The licensee had sufficient time to adjust the individual parameters used in the 3-SR-3.1.7.6 calculation so that normal pump flow deviation that occurred on October 18, 2023, would not have led to both SLC subsystem being declared inoperable.

Enforcement:

Violation: BFN Unit 3 TS 5.4.1.a states in part that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2, Appendix A, Section 4.d recommends, in part, procedures for energizing, filling, venting, and draining for the standby liquid control system.

Contrary to the above, from March 23, 2018 until October 18, 2022, the licensee failed to establish, implement, and maintain procedures for energizing, filling, venting, and draining the standby liquid control system. Specifically, the licensee failed to maintain the Unit 3 standby liquid control storage tank filled with a sufficient concentration of sodium pentaborate to maintain compliance with TS SR 3.1.7.6.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Pressure Boundary Leak on Shutdown Cooling Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000296/2023003-05 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 3 Technical Specification 3.4.4 and 10 CFR 50, Appendix B, Criterion III, "Design Control,"

was identified when the licensee failed to accurately model the supports needed for small bore piping associated with the Unit 3 loop I residual heat removal shutdown cooling line. As a result, a pressure boundary leak developed on a vent line for this shutdown cooling line when flow induced vibrations caused cyclic fatigue failure at the location.

Description:

On December 3, 2022, during a drywell walkdown for Unit 3 forced outage F311 to identify the cause for increased drywell leakage, BFN personnel discovered a through-wall piping leak on a vent pipe assembly for the loop I residual heat removal (RHR) shutdown cooling line. This assembly was classified as ASME Code Class 1 piping and constituted part of the Unit 3 reactor coolant system pressure boundary.

The vent pipe assembly was a 3/4 schedule 80 pipe welded to the top of the residual heat removal shutdown cooling line. This leak was determined to be unisolable from the reactor pressure vessel. The licensee made an event notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded - and submitted LER 05000296/2022-003 (ADAMS Ascension number ML23031A357) on January 31, 2023.

This condition is assumed to have developed on August 6, 2022, when system monitoring detected an increase in unidentified drywell leakage. The licensee modified the vent line by cutting the vent line, removing the vent valves, and capping the vent line. The licensee determined that the leak occurred in a socket weld connection at one of the vent valves in the line caused by high cycle fatigue that exceeded the endurance limit.

During evaluations performed for extended power uprate in 2007, the small bore piping was considered rigid and, therefore, did not need to be modeled independently from the large bore piping. Based on how this line was modeled, the licensees fatigue evaluation concluded that infinite vibration cycles were applicable. However, when modeled independently with an accurate fatigue stress input, infinite vibration cycles no longer apply. Consequently, design calculation CDQ106820020029, Pipe Stress Analysis of Pipe Stress Problem No. N1-168-1R, failed to identify that the stress caused by flow induced vibration at normal operating conditions was sufficient to exceed the endurance limit for the failed pipe assembly.

Corrective Actions: A temporary modification was implemented to cut and cap the vent line until addressed during the next Unit 3 outage. 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 smallbore piping with vulnerability to fatigue failure due to exceeding the endurance limit due to operational vibration.

Corrective Action References: CR 1820718 and 1747875

Performance Assessment:

Performance Deficiency: The failure to accurately model the small bore piping in the residual heat removal system, as required by 10CFR 50, Appendix B, Criterion III, Design Control, was a performance deficiency. Specifically, the licensee mischaracterized the rigidity of the failed pipe assembly and failed to accurately calculate the endurance limit of the failed pipe assembly. This failure ultimately led to a pressure boundary leak which could have been predicted and prevented with accurate modeling.

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 licensee mischaracterized the rigidity of the failed pipe assembly and failed to accurately calculate the endurance limit of the failed pipe assembly. This failure ultimately led to a pressure boundary leak which could have been predicted and prevented with accurate modeling.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened out in the review of the Barrier Integrity cornerstone as the performance deficiency was not related to pressurized thermal shock; therefore, the finding was addressed under the Initiating Events cornerstone. Since a reasonable assessment of degradation, could have resulted in exceeding the reactor coolant system leak rate for a small loss of coolant accident, a detailed risk evaluation was performed by a regional Senior Reactor Analyst in accordance with IMC 0609, Appendix A, utilizing the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations 8 Version 8.2.8, and the NRC Browns Ferry Unit 3 Standardized Plant Analysis Risk model version 8.80 dated 3/31/2022. The exposure period was from August 6, 2022, when indications of the leak were present until December 3, 2022, when the plant was taken to mode 4, a total of 119 days. The performance deficiency was conservatively modelled as an increase in the small loss of coolant accident frequency by two orders of magnitude given the leak of the vent line had the potential to cause the socket to fail and initiate an unisolable small break loss of coolant accident if the entire socket failed. The dominant sequence was a small break loss of coolant accident with failure of the turbine bypass valves, failure of high-pressure injection and operators failing to manually depressurize the reactor. The detailed risk evaluation estimated that the performance deficiency resulted in an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: BFN Unit 3 TS LCO 3.4.4 requires, in part, that operational leakage shall be limited to no pressure boundary leakage while in Modes 1, 2, and 3; otherwise, the unit shall be shut down and 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 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Contrary to the above, from approximately August 6, 2022, to December 3, 2022, Unit 3 operational leakage included pressure boundary leakage while in Mode 1, the unit was not shut down and 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 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

10 CFR 50, Appendix B, Criterion III, "Design Control," requires in part that design control measures shall be established to provide for verifying or checking the adequacy of design.

Contrary to the above, from approximately 2007 until December 3, 2022, the licensee failed to establish design control measures to verify or checking the adequacy of design of the supports needed for the small bore piping associated with the Unit 3 loop I residual heat removal shutdown cooling line.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Pressure Boundary Leak on Reactor Recirculation Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000260/2023003-04 Open/Closed A self-revealed Green finding and associated non-cited violation of Browns Ferry Nuclear Unit 2 Technical Specification 3.4.4 and 10 CFR 50, Appendix B, Criterion III, "Design Control,"

was identified when the licensee failed to accurately model the supports needed for the drain pipe assembly for the Unit 2 recirculation pump A discharge valve, 2-FCV-068-0003. As a result, a pressure boundary leak developed when flow induced vibrations caused cyclic fatigue failure at the location.

Description:

On February 18, 2023, during an initial drywell walkdown for refueling outage

2R22 when the reactor was in Mode 4, BFN personnel discovered a 15 drop per minute

through-wall piping leak on the drain pipe assembly for the recirculation pump A discharge valve, 2-FCV-068-0003. This assembly was classified as ASME Code Class 1 piping and constituted part of the Unit 2 reactor coolant system pressure boundary.

The drain pipe assembly was a 3/4 schedule 80 pipe welded to the drain hole in the bottom of the valve body. This leak was determined to be unisolable from the reactor pressure vessel.

The licensee made an event notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded - and submitted LER 05000260/2023-001 (ADAMS Ascension number ML23109A362) on April 19, 2023.

This condition is assumed to have developed in November of 2022, when system monitoring detected an increase in unidentified drywell leakage. The licensee removed the drain line and welded a 2 pipe stub to the bottom of the valve body to act as the valve body pressure boundary. The licensee determined that the leak occurred in a socket weld connection between the pipe nipple and the valve body caused by high cycle fatigue that exceeded the endurance limit.

During evaluations performed for extended power uprate in 2007, the small bore piping was considered rigid and, therefore, did not need to be modeled independently from the large bore piping. Based on how this line was modeled, the licensees fatigue evaluation concluded that infinite vibration cycles were applicable. However, when modeled independently with an accurate fatigue stress input, infinite vibration cycles no longer apply. Consequently, design calculation CDQ106820020029, Pipe Stress Analysis of Pipe Stress Problem No. N1-168-1R, failed to identify that the stress caused by flow induced vibration at normal operating conditions was sufficient to exceed the endurance limit for the failed pipe assembly.

Corrective Actions: A temporary modification was implemented to remove the test valves and vent piping to resolve leakage until addressed during the next Unit 2 outage. 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.

Corrective Action References: CR 1836572 and 1747875

Performance Assessment:

Performance Deficiency: The failure to accurately model the drain pipe assembly for the 2-FCV-068-0003, as required by 10CFR50, Appendix B, Criterion III was a performance deficiency. Specifically, the licensee mischaracterized the rigidity of the failed pipe assembly and failed to accurately calculate the endurance limit of the failed pipe assembly. This failure ultimately led to a pressure boundary leak which could have been predicted and prevented with accurate modeling.

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 inadequate application of corrective actions from previous fatigue failures of ASME Code Class 1 equivalent socket welded connections resulted in an un-isolable through wall leak in the reactor coolant system.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened out in the review of the Barrier Integrity cornerstone as the performance deficiency was not related to pressurized thermal shock; therefore, the finding was addressed under the Initiating Events cornerstone. Since a reasonable assessment of degradation, could have resulted in exceeding the reactor coolant system leak rate for a small loss of coolant accident, a detailed risk evaluation was performed by a regional Senior Reactor Analyst in accordance with IMC 0609, Appendix A, utilizing the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations 8 Version 8.2.6, and the NRC Browns Ferry Unit 2 Standardized Plant Analysis Risk model version 8.80 dated 5/26/2022. The exposure period was from November 13, 2022, when indications of the leak were present until February 17, 2023, when the plant was taken to mode 4, a total of 97 days. The performance deficiency was conservatively modelled as an increase in the small loss of coolant accident frequency by two orders of magnitude given the leak of 2-FCV-068-0003 had the potential to cause the socket to fail and initiate an unisolable small break loss of coolant accident if the entire socket failed. The dominant sequence was a small break loss of coolant accident with failure of the turbine bypass valves, failure of high-pressure injection and operators failing to manually depressurize the reactor. The detailed risk evaluation estimated that the performance deficiency resulted in an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: BFN Unit 2 TS LCO 3.4.4 requires, in part, that operational leakage shall be limited to no pressure boundary leakage while in Modes 1, 2, and 3; otherwise, the unit shall be shut down and 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 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Contrary to the above, from approximately November 13, 2022, to February 17, 2023, Unit 2 operational leakage included pressure boundary leakage while in Mode 1, the unit was not shut down and 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 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

10 CFR 50, Appendix B, Criterion III, "Design Control," requires in part that design control measures shall be established to provide for verifying or checking the adequacy of design.

Contrary to the above, from 2007 until February 17, 2023, the licensee failed to establish design control measures to verify or checking the adequacy of design of the supports needed for the drain pipe assembly for the Unit 2 recirculation pump A discharge valve, 2-FCV-068-0003.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Browns Ferry Nuclear Plant, Unit 1, Main Steam Relief Valves Lift Settings Outside of Technical Specification Required Setpoints Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000259/2023003-06 Applicable Open/Closed A self-revealed Severity Level IV non-cited violation of Browns Ferry Nuclear Unit 1 Technical Specifications 3.4.3 and LCO 3.0.4 was identified when the licensee discovered, through as found test results, that two of thirteen main steam relief valves that were removed for testing had as found lift settings outside of the +/-3 percent setpoint band required for their operability.

Description:

The BFN Unit 1 TS 3.4.3 requires twelve of the thirteen main steam relief valves (MSRVs) to be operable while in Modes 1, 2, and 3. On December 6, 2022, the Tennessee Valley Authority was notified of as-found testing results that two MSRVs from Unit 1 were outside of the +/-3 percent setpoint band required for operability. It was determined that these MSRVs failed due to corrosion bonding between the pilot valve disc and seat and simmering. Both MSRVs were considered to be inoperable during the entire operating cycle from January 23, 2022, to September 30, 2022, 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 January 23, 2022, BFN Unit 1 entered a TS 3.4.3 Applicable Mode when LCO TS 3.4.3 Required Actions were not met. Therefore, Unit 1 was also in violation of TS 3.0.4.

The affected valves remained capable of maintaining reactor pressure below the American Society of Mechanical Engineers code limit of 1375 psig. All thirteen of the MSRV pilot valves were replaced during the Unit 1 fall 2020 refueling outage. The previous corrective action from LER 05000260/2019-002-00 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, had not yet been implemented for these valves.

Corrective Actions: The licensee replaced all thirteen MSRV pilot valves during the fall 2022 refueling outage. 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 +/- one 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, and 1822254.

Performance Assessment:

The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.

Enforcement:

Severity: This violation is characterized as a Severity Level IV NCV based on its similarity to the SL IV examples in Section 6.1.d 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: BFN Unit 1 TS 3.4.3.A requires that with one or more required MSRVs inoperable, that 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, two required MSRVs were inoperable from November 6, 2020, to September 30, 2022, 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.

BFN Unit 1 TS 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 January 23, 2023, Browns Ferry Nuclear Unit 1 entered a TS 3.4.3 applicable mode when TS 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 October 5, 2023, the inspectors presented the integrated inspection results to Mr.

Darrell Lock, Assistant Plant Manager, and other members of the licensee staff.

• On August 9, 2023, the inspectors presented the Radiation Protection Debrief Meeting inspection results to Tony Hairston, Radiation Protection Manager and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Drawings 1-47E811-1 Flow Diagram Residual Heat Removal System Rev 48

Procedures 0-OI-18 Fuel Oil System Rev 58

1-OI-74 Residual Heat Removal System Rev 125

1-OI-75, Attachment Core Spray System Valve Lineup Checklist 21

3-OI-82 Standby Diesel Generator System Rev 157

Work Orders WO 123030205

71111.05 Corrective Action CR 1879887

Documents

Fire Plans FPR-Volume 2 Fire Protection Report Volume 2 Rev 77

71111.07A Work Orders 123068155

71111.12 Miscellaneous Maintenance rule 12th periodic report 03/10/20

Maintenance rule 13th periodic report 03/10/22

Tennessee Valley Authority approved suppliers list excerpt

for Engine Systems, Inc.

6947661 Purchase order for new heat exchangers for 3A 0

Emergency Diesel Generator

6947661 Material inspection form for replacement 3A emergency 03/02/2022

diesel generator heat exchangers

71111.13 Corrective Action 1871312 07/30/2023

Documents

Procedures 0-OI-31 Control Bay and Off-Gas Treatment Building Air Rev 174

Conditioning System

71111.15 Corrective Action Condition Reports 1872763, 1879889, 1879991, 1879992, 1873186,

Documents 1875761, 1878492, 1880541, 1880309, 1880021, 1883072

Miscellaneous MDQ 0031970069 Rev 9

Control Bay Chiller Adequacy Analysis

Procedures 0-SR-3.3.7.1.3(B) Control Room Air Supply Duct Radiation Monitor, 0-RM-90- Rev 37

259B, Calibration and Functional Test

0-TI-158 0-TI-158, Representative/Bottom Sampling of the Diesel 32

Generator 7-Day Tank Fuel-Oil performed on 8/4/23

0-TI-158 0-TI-158, Representative/Bottom Sampling of the Diesel 32

Inspection Type Designation Description or Title Revision or

Procedure Date

Generator 7-Day Tank Fuel-Oil performed on 8/5/23

3-OI-73 High Pressure Coolant Injection System Rev 67

Work Orders Work Orders 121597600, 123169346

71111.18 Calculations MDQ007320100031 High Pressure Coolant Injection Pump and System 2

Hydraulic Analysis

Engineering 71866 Modify 1-FCV-073-0035 to meet stroke time requirements 12/15/2015

Changes 71867 Modify 2-FCV-073-0035 to meet stroke time requirements 12/15/2015

71868 Modify 3-FCV-073-0035 to meet stroke time requirements 11/30/2015

TMOD BFN-2-2023- Cut and cap test line for residual heat removal test valves, 0

074-001

Work Orders 123416903

71111.20 Engineering 71867 Modify 2-FCV-073-0035 to meet stroke time requirements 12/15/2015

Changes

71111.24 Corrective Action Condition Report 1872512, 1879889

Documents

Miscellaneous M&TE certificate of For torque screwdriver E48212 9/12/2022

calibration 247458

M&TE certificate of For torque wrench E51500 10/26/2022

calibration 249754

M&TE certificate of For torque wrench E60592 2/13/2023

calibration 255940

Procedures 1-SI-4.4.A.1 Standby Liquid Control Pump Functional Test

1-SR-3.1.7.3 Standby Liquid Control System Enriched Sodium 56

Pentaborate Solution Concentration, Quantity Calculation,

and ATWS Equivalency Calculation

3-SR-3.8.1.1(3A) Diesel Generator 3A Monthly Operability Test Rev 72

Work Orders Work Orders 120151834, 121469373, 121471563, 121597617,

23052369, 123071652, 123153907, 123197462,

23169346, 121597600, 123030205, 119334041,

2396777, 123808713, 123808788, 123971610

71114.06 Miscellaneous 2023 Browns Ferry Nuclear August Training Drill Scenario 08/02/2023

and Controller Book

71124.01 Corrective Action CR 1870990

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents

Radiation M-20230727-10 U2 RXB 621 Fuel Pool Cooling Area 07/27/2023

Surveys

Radiation Work RWP 23220552 Unit 2 Operations Activities Revision 1

Permits (RWPs)

71152A Corrective Action Condition Reports 1747875, 1820718

Documents

71153 Corrective Action CR 1830955

Documents

28