IR 05000259/2024002

Brown Ferry Nuclear Plant – Integrated Inspection Report05000259/2024002 and 05000260/2024002 and 05000296/2024002

ML24214A167
Person / Time
Site:
Issue date:	08/02/2024
From:	Louis Mckown Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
IR 2024002
Download: ML24214A167 (1)

Text

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2024002 AND 05000260/2024002 AND 05000296/2024002

Dear Jim Barstow:

On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On July 9, 2024, the NRC inspectors discussed the results of this inspection with E. Quinn Leonard, Plant Manager 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. Two Severity Level IV violations without an associated finding are 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 and Severity Level IV is documented in this report. We are treating this violation as a non-cited violation (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.

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.August 2, 2024

Sincerely, Louis J. McKown, II, Chief Reactor Projects Branch 5 Division of Reactor Projects 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: 05000259/2024002, 05000260/2024002 and 05000296/2024002

Licensee: Tennessee Valley Authority

Facility: Browns Ferry Nuclear Plant

Location: Athens, Alabama

Inspection Dates: April 01, 2024 to June 30, 2024

Inspectors: S. Billups, Resident Inspector K. Pfeil, Resident Inspector T. Steadham, Senior Resident Inspector J. Tornow, Physical Security Inspector J. Walker, Senior Emergency Preparedness Inspector

Approved By: Louis 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. A licensee-identified non-cited violation is documented in report section: 7115

List of Findings and Violations

Non-functional Wide Range Gaseous Effluent Radiation Monitor Rendering EALs Ineffective Cornerstone Significance Cross-Cutting Report Aspect Section Emergency Green [P.3] - 71114.05 Preparedness NCV Resolution 05000259,05000260,05000296/202400 2-01 Open/Closed The inspectors identified a Green Finding and associated Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.47(b)(4) for unavailable indication of the wide range noble gas monitor, which rendered several emergency action levels (EALs)ineffective.

Failure to meet Technical Specifications for Degraded Damper Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259,05000260,05000296/202400 2-02 Open/Closed A self-revealed Severity Level IV NCV of Technical Specification (TS) 3.6.4.2 was identified when the reactor zone outboard supply damper, a TS-required secondary containment isolation valve, failed to close after a valid signal to close. A past operability evaluation determined that the damper had failed approximately three months prior which constituted a failure to meet the TS requirements for secondary containment isolation valve operability because it was not identified to be inoperable at the time.

Inoperability of Unit 3 Diesel Generator due to Relay Failure Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259,05000260,05000296/202400 2-03 Open/Closed A self-revealed Severity Level IV NCV of Technical Specification (TS) 3.8.1, 3.8.2 and TS 3.0.4 was identified when the 3D standby diesel generator (SDG) failed to start during testing due to a failed relay. Specifically, a past operability evaluation determined that the 3D SDG was inoperable between December 14, 2022, until February 15, 2024.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000259,05000260,05 LER 2024-001-00 for 71153 Closed 000296/2024-001-00 Browns Ferry Nuclear Plant,

Units 1, 2, and 3,

Inoperability of Unit 3 Diesel Generator due to Relay Failure

LER 05000259/2024-002-00 LER 2024-002-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 1, Reactor Scram due to Generator Step-Up Transformer Failure

LER 05000260/2024-001-00 LER 2024-001-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 2, Secondary Containment Isolation Valve Inoperable due to Mechanical Failure

LER 05000260/2024-001-01 LER 2024-001-01 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 2, Secondary Containment Isolation Valve Inoperable due to Mechanical Failure

LER 05000296/2024-001-00 LER 2024-001-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 3, Primary Containment Isolation Valve Inoperable due to Incorrect Motor Operated Valve Setup

LER 05000296/2024-002-00 LER 2024-002-00 for 71153 Closed Browns Ferry Nuclear Plant,

Unit 3, Breaker Trip Automatically Started an Emergency Diesel Generator

PLANT STATUS

Unit 1 began the inspection period at full (100 percent) rated thermal power (RTP). On April 19, 2024, operators lowered reactor power to 60 percent RTP for a control rod sequence exchange. On April 20, 2024, the unit was returned to 100 percent RTP. On April 24, 2024, a scram occurred causing forced outage F112 due to a loss of the B phase of the main generator output transformer. After replacing the transformer, operators restarted the unit on May 5, 2024. On May 8, 2024, the unit was returned to 100 percent RTP. On June 14, 2024, operators lowered reactor power to 70 percent RTP for a control rod sequence exchange. On June 15, 2024, 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 May 23, 2024, operators lowered reactor power to 60 percent RTP for a control rod sequence exchange. On May 25, 2024, the unit was returned to 100 percent RTP.

On June 26, 2024, operators lowered reactor power to 70 percent RTP for main condenser water box cleaning. On June 28, 2024, 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 May 17, 2024, operators lowered reactor power to 57 percent RTP for a control rod sequence exchange. On May 18, 2024, the unit was returned to 100 percent RTP. On June 7, 2024, operators lowered reactor power to 57 percent RTP for power suppression testing to identify and suppress a suspected leaking fuel assembly. On June 10, 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 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) (7 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Unit 1 C standby diesel generator walkdown after planned maintenance outage on April 23, 2024

(2) Unit 3 reactor core isolation cooling system after comprehensive pump test on May 03, 2024

(3) Unit 1 standby liquid control after functional test on May 14, 2024

(4) Unit 1 emergency high pressure makeup system on May 14, 2024

(5) Unit 2 residual heat removal loops 1&2 after maintenance on May 16, 2024

(6) Unit 2 high pressure coolant injection after maintenance on May 16, 2024

(7) Unit 3 control bay chiller 3A while Unit 3 control bay chiller 3B was out of service for maintenance on May 17, 2024

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (8 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 3 cable spreading room on April 09, 2024

(2) Unit 1 & 2, diesel generator building 565' elevation on April 17, 2024

(3) Unit 3, fire area 3-1 NW corner room, reactor building elevation 519' & 541' on May 02, 2024

(4) Unit common, refuel floor on May 09, 2024

(5) Unit common, 1C hallway on May 10, 2024

(6) Unit 1, fire area 1-5, reactor building elevation 639' on May 14, 2024

(7) Unit 1, turbine building, elevation 557' & 565', emergency high pressure makeup on May 14, 2024

(8) Unit 2, fire areas 2-1 & 2-2, SW & SE corner rooms elevation 519' & 541' on May 15, 2024

71111.07A - Heat Exchanger/Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Unit 1 C diesel generator cooling water heat exchanger C1 and C2 inspection under WO 123420296 on April 17, 2024

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)

(1) Operator performance during Unit 1 main control room after forced outage F112, automatic plant scram and putting shutdown cooling in service after 1b main transformer failure on April 25, 2024

(2) Operator performance during unit 3 main control room power ascension, reactivity maneuvering, & single rod Scram 38-31 on June 10, 2024

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) Maintenance risk during Unit 1 preferred transformer maintenance on April 18, 2024

(2) Maintenance risk during diesel generator C maintenance outage on April 19, 2024

(3) Emergent work control for Unit 1 residual heat removal shutdown cooling secured on April 26, 2024

(4) Emergent work control for Unit 1 4kv shutdown boards on alternate power feed on April 26, 2024

(5) Emergent work control during Unit 2 loop I core spray maintenance outage on May 09, 2024

(6) Maintenance risk during D2 residual heat removal service water maintenance on May 15, 2024

(7) Maintenance risk during 3B control bay chiller out of service for planned maintenance on May 16, 2024

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)

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

(1) CR 1902390, standby diesel generator 3C electrical panel doors not fully latched on April 16, 2024

(2) CR 1926099, standby diesel generator 3D output breaker failed to manually close on May 02, 2024

(3) CR 1938714, as-found overspeed trip setpoint out of tolerance for D standby diesel generator on June 27, 2024

(4) Unit 3 high pressure coolant injection rupture disc extent of condition as a result of failed rupture disc on Unit 2 on June 27, 2024

(5) CR 1940354, residual heat removal pump 3B did not meet pump performance acceptance criteria on June 28, 2024

(6) CR 1940394, elevated vibrations on Unit 2 high pressure coolant injection pump on June 28, 2024

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) Unit 2 long term scaffold storage area modification on May 28, 2024

(2) BFN-24-012, disconnect torque switch and torque switch bypass from main steam line drain valve inboard containment isolation valve, BFN-3-FCV-001-0055, close circuit on June 25, 2024

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated various Unit 1 forced outage F112 activities from April 25, 2024, through May 6, 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) (5 Samples)

(1) WO 122901749, Unit 3 operation standby liquid control preventative maintenance test after squib valve replacement on April 15, 2024

(2) WO 123405588, C diesel generator 2-year inspection/pm on April 29, 2024

(3) WO 124446965, replace Unit 1 high pressure coolant injection rupture discs on May 28, 2024

(4) WO 123785860, perform battery replacement shutdown board D battery on June 17, 2024

(5) WO 117608911, replace the generator for the D diesel generator on June 21, 2024

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) WO 123511382, C diesel generator monthly operability test following planned maintenance, procedure 0-SR-3.8.1.1(C) on April 23, 2024

(2) WO 123543722, Unit 3 reactor core isolation cooling comprehensive pump test SR 3.5.3.1 & 3.5.3.3 on May 03, 2024

(3) WO 123585976, Unit 1 standby liquid control pump functional test, 1-SI-.4.A.1 on May 20, 2024

71114.02 - Alert and Notification System Testing

Inspection Review (IP Section 02.01-02.04) (1 Sample)

(1) The inspectors evaluated the maintenance and testing of the alert and notification system during the week of April 8, 2024.

71114.03 - Emergency Response Organization Staffing and Augmentation System

Inspection Review (IP Section 02.01-02.02) (1 Sample)

(1) The inspectors evaluated the readiness of the Emergency Response Organization (ERO) during the week of April 8, 2024.

71114.04 - Emergency Action Level and Emergency Plan Changes

Inspection Review (IP Section 02.01-02.03) (1 Sample)

(1) The inspectors evaluated submitted Emergency Action Level (EALs), Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of April 8, 2024. This evaluation does not constitute NRC approval.

71114.05 - Maintenance of Emergency Preparedness

Inspection Review (IP Section 02.01 - 02.11) (1 Sample)

(1) The inspectors evaluated the maintenance of the emergency preparedness program during the week of April 8, 2024.

71114.06 - Drill Evaluation

Required Emergency Preparedness Drill (1 Sample)

The inspectors observed and evaluated an emergency preparedness drill which formally assessed elements contained within NEI 99-02 for performance indicators regarding the Emergency Preparedness Cornerstone sections pertaining to Drill Exercise Performance (DEP).

(1) The inspectors observed and evaluated an emergency preparedness drill on April 24, 2024. Events included an anticipated transient without scram, loss of fuel clad/reactor coolant system barrier, and loss of reactor pressure vessel level indication.

Additional Drill and/or Training Evolution (1 Sample)

The inspectors evaluated:

(1) The inspectors observed and evaluated an emergency preparedness drill on June 26, 2024. Events included a control rod drift leading to a manual scram, a reactor coolant system leak in the drywell, and an unisolable leak outside the drywell.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

(3) Unit 3 (April 1, 2023, through March 31, 2024)

MS07: High Pressure Injection Systems (IP Section 02.06) (3 Samples)

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

(3) Unit 3 (April 1, 2023, through March 31, 2024)

MS08: Heat Removal Systems (IP Section 02.07) (3 Samples)

(1) Unit 1 (April 1, 2023, through March 31, 2024)

(2) Unit 2 (April 1, 2023, through March 31, 2024)

(3) Unit 3 (April 1, 2023, through March 31, 2024)

EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)

(1) April 1 through Dec 31, 2023.

EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)

(1) April 1 through Dec 31, 2023.

EP03: Alert And Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)

(1) April 1 through Dec 31, 2023.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Follow up (IP Section 03.01)

(1) Response and event follow-up to Unit 1 unplanned scram on April 25, 2024

Event Report (IP Section 03.02) (6 Samples)

The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.

(1) LER 2024-001-000 for Browns Ferry Nuclear Plant, Unit 1, Inoperability of Unit 3 Diesel Generator due to Relay Failure (ADAMS Accession No. ML24102A279). The inspectors determined that this LER was related to a SL IV NCV that is documented in the violations section of this report and a performance deficiency that was previously dispositioned as NCV 05000296/2024001-01 (ADAMS Accession No.

ML24131A141.) This LER is Closed.

(2) LER 2024-002-00 for Browns Ferry Nuclear Plant, Unit 1, Reactor Scram due to Generator Step-Up Transformer Failure (ADAMS Accession No. ML24176A102). The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified. No violations were identified as a result of this review. This LER is Closed.

(3) LER 2024-001-000 for Browns Ferry Nuclear Plant, Unit 2, Secondary Containment Isolation Valve Inoperable due to Mechanical Failure (ADAMS Accession No. ML24047A222). The inspectors determined that this LER was related to a SL IV NCV violation that is documented in the violations section of this report. This LER is Closed.

(4) LER 2024-001-001 for Browns Ferry Nuclear Plant, Unit 2, Secondary Containment Isolation Valve Inoperable due to Mechanical Failure (ADAMS Accession No. ML24108A183). The inspectors determined that this LER was a revision to an LER previously reviewed in this inspection report. No additional violations were identified as a result of this review. This LER is Closed.

(5) LER 2024-001-00 for Browns Ferry Nuclear Plant, Unit 3, Primary Containment Isolation Valve Inoperable due to Incorrect Motor Operated Vale Setup (ADAMS Accession No. ML24113A201). The inspectors determined that this LER was related to a SL IV NCV violation that is documented in the violations section of this report. This LER is Closed.

(6) LER 2024-002-00 for Browns Ferry Nuclear Plant, Unit 3, Breaker Trip Automatically Started an Emergency Diesel Generator (ADAMS Accession No.

ML24115A165). The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified. No violations were identified as a result of this review. This LER is Closed.

INSPECTION RESULTS

Non-functional Wide Range Gaseous Effluent Radiation Monitor Rendering EALs Ineffective Cornerstone Significance Cross-Cutting Report Aspect Section Emergency Green [P.3] - 71114.05 Preparedness NCV Resolution 05000259,05000260,05000296/20240 02-01 Open/Closed The inspectors identified a Green Finding and associated Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.47(b)(4) for unavailable indication of the wide range noble gas monitor, which rendered several emergency action levels (EALs)ineffective.

Description:

During an emergency facility walkdown on April 11, 2024, the inspectors identified the plant wide range gaseous effluent radiation monitor (WRGERMS) in the Unit 1

& 2 control room was out of service. The inspectors queried the licensee to determine if compensatory equipment to provide WRGERMS data to the operators was available, but the licensee indicated there was none. At the time of inspection, the WRGERMS monitor had been out of service since June 20, 2023. The absence of WRGERMS indication effected the licensees ability to declare EALs within the Radiological Initiation conditions for RG1, RS1, RA1, and RU1, and the ability to perform dose assessments.

Corrective Actions: A standing order was put in place directing plant operations to use alternate stack radiation indication in the control room to monitor rising radiation levels. If necessary, the stack release rates are calculated utilizing manual sampling from existing procedures. In the event that the alternate monitors to WRGERMS were to become inoperable (WRGERMS and alternate radiation monitor indication are inoperable), another existing backup procedure is available to obtain stack release rate information and allow the operators to obtain information necessary to determine EAL thresholds.

Corrective Action References: CR 1923873

Performance Assessment:

Performance Deficiency: Failure to provide indication of the wide range noble gas monitor in the Unit 1 & 2 control room, which rendered several emergency action levels (EALs)ineffective.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the ERO Performance attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors determined the performance deficiency was more than minor because it was associated with the Emergency Response Organization (ERO) Performance attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix B, Emergency Preparedness SDP. The inspectors assessed the significance of the finding using IMC 0609 Appendix B, Emergency Preparedness SDP. Utilizing Table 5.4-1, "Significance Examples of §50.47(b)(4)," and Figure 5.4-1, "Significance Determination for Ineffective EALs and Overclassification," the NRC determined that an appropriate declaration could still be made in an accurate and timely manner using the alternate method to determine plant stack release rates. This represents a finding of very low safety significance (Green).

Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance.

Enforcement:

Violation: 10 CFR 50.47(b)(4) requires in part a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee."

10 CFR 50.47(b)(9) requires in part that adequate equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.

Contrary to the above, from June 20, 2023 until April 11, 2024, the licensee failed to provide continuous indication of plant wide range gaseous effluent radiation levels in the control room to plant operators, rendering EALs related to determining radiation levels using that parameter ineffective.

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

Failure to meet Technical Specifications for Degraded Damper Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV Applicable 05000259,05000260,05000296/2024002-02 Open/Closed A self-revealed Severity Level IV NCV of Technical Specification (TS) 3.6.4.2 was identified when the reactor zone outboard supply damper, a TS-required secondary containment isolation valve, failed to close after a valid signal to close. A past operability evaluation determined that the damper had failed approximately three months prior which constituted a failure to meet the TS requirements for secondary containment isolation valve operability because it was not identified to be inoperable at the time.

Description:

The Browns Ferry Units 1, 2, and 3 TS 3.6.4.2 requires each secondary containment isolation valve to be operable while in Modes 1, 2, and 3. With one valve inoperable, Condition A requires the other valve in the penetration to be closed and deactivated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If Condition B is not met, Condition C requires 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 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance requirement 3.6.4.2.2 requires that each isolation valve actuates to the isolation position on an actual or simulated actuation signal.

On August 20, 2023, while performing surveillance procedure 0-SR-3.6.4.2.1, Secondary Containment Isolation Valve Stroke Timing, the reactor zone outboard supply damper, 2-DMP-064-0013, failed to close on an actual isolation signal when the reactor zone supply fans were secured. The licensee declared 2-DMP-064-0013 inoperable and troubleshooting identified no deficiencies. The licensee concluded that this failure was an isolated intermittent failure, successfully reperformed the surveillance test, and declared the damper operable.

On November 18, 2023, while performing 0-SR-3.6.4.2.1, 2-DMP-064-0013 failed to close in the same manner as it did on August 20, 2023. Troubleshooting identified wear on the damper louver arm pin holes which allowed the louver arm to become loose and ultimately jam on top of the louver bracket. On November 22, 2023, operations declared the damper operable following corrective maintenance and testing.

The licensee later determined that the failure of the damper to close on August 20, 2023, was also attributed to this wear on the damper louver arm pin holes. The licensee therefore concluded that 2-DMP-064-0013 was inoperable since August 20, 2023, which was longer than allowed by TS. This condition did not result in a loss of safety function because the Reactor Zone Inboard Supply Damper remained operable. The licensee submitted Licensee Event Report 50-260/2024-001-01 to document the operation prohibited by the plants TS.

The inspectors reviewed the previous failures of this damper and the corrective actions associated with the troubleshooting. The inspectors also reviewed the procedure that was used to perform preventative maintenance on this damper. Because there was no conclusive evidence that the dampers were degraded in March 2021, when the damper was last inspected, the inspectors did not consider the enhancements made to the procedure to represent a performance deficiency. The inspectors concluded that because the troubleshooting efforts were reasonable given the information available to the licensee at the time, the cause of the failure was not reasonably within the licensees ability to foresee and correct.

Corrective Actions: The licensee replaced all broken and/or degraded components on 2-DMP-064-0013 and revised applicable maintenance procedures to include enhanced dampener preventative maintenance guidance recommended by the Electric Power Research Institute.

Corrective Action References: CRs 1882682, 1883221, 1893273, 1893751, and 1893994

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 SLIV 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 Plant, Units 1, 2, and 3 TS 3.6.4.2, Secondary Containment Isolation Valves, Condition A, requires that with one or more penetration flow paths with one SCIV inoperable, Action A.1 - Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. TS 3.6.4.2 Condition C requires that with the required action and associated Completion time of Condition A or B not met, that the units 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, from August 20, 2023, to November 22, 2023, with one secondary containment isolation valve were inoperable, the affected penetration flow path was not isolated by the use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and the units 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. Specifically, 2-DMP-064-0013, a secondary containment isolation valve, was inoperable starting 12:15 a.m. on August 20, 2023, and remained inoperable until repairs were completed on November 22, 2023. During this time, the inboard damper in this secondary containment penetration was neither closed nor de-activated.

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

Inoperability of Unit 3 Diesel Generator due to Relay Failure Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV Applicable 05000259,05000260,05000296/2024002-03 Open/Closed A self-revealed Severity Level IV NCV of Technical Specification (TS) 3.8.1, 3.8.2 and TS 3.0.4 was identified when the 3D standby diesel generator (SDG) failed to start during testing due to a failed relay. Specifically, a past operability evaluation determined that the 3D SDG was inoperable between December 14, 2022, until February 15, 2024.

Description:

As described in inspection report 05000296/2024001 (ADAMS Accession Number ML24131A141), the inspectors previously identified a finding associated with the February 14, 2024, failure of the 3D SDG. On April 11, 2024, the licensee submitted LER 00259/2024-001 to document the TS violations associated with the inoperability of the 3D SDG from December 28, 2022, until February 15, 2024.

Units 1, 2, and 3 TS 3.8.1, Condition I requires, in part, that for one Unit 3 SDG inoperable for more than 14 days to enter Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Because the 3D SDG was inoperable since 8:00 a.m. on December 14, 2022, the inspectors concluded that the failure to comply with TSs started for all three Units on 8:00 p.m. on December 28, 2022.

Units 1, 2, and 3 TS Section 3.0, 'LCO Applicability', Subsection 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. None of the Units TS allow for unlimited operation with an SDG inoperable. The inspectors reviewed the Mode changes from Mode 4 to Mode 3 for each unit and determined that each Unit made an inappropriate Mode change on the following days:

• February 1, 2023, May 22, 2023, and January 30, 2024, for Unit 1

• March 22, 2023, for Unit 2

• April 28, 2023, for Unit 3

The inspectors reviewed the LER and identified no additional performance deficiencies. However, as described in the LER, the inoperability of the 3D SDG for this time frame represented the operation of all three units in a condition prohibited by TS.

Corrective Actions: The licensee replaced the failed relay.

Corrective Action References: CRs 1823305 and 1910087.

Performance Assessment:

The NRC determined that this violation was associated with a previously documented finding assessed using the significance determination process.

Enforcement:

Severity: This violation is characterized as a Severity Level IV NCV based on its similarity to SLIV 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 previously dispositioned performance deficiency screened as Green for an exposure time of 1 year.

Violation: Browns Ferry Nuclear Plant, Units 1, 2, and 3 TS Subsection 3.8.1, AC Sources -

Operating, Condition I, requires that with one Unit 3 SDG 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, from December 28, 2022, to February 15, 2024, one Unit 3 SDG was inoperable for greater than 14 days and the units 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 Plant, Units 1, 2, and 3 TS Subsection 3.8.2, AC Sources -

Shutdown, Condition B, requires that with one Unit 3 SDG inoperable to immediately suspend core alterations, immediately suspend movement of recently irradiated fuel assemblies in the secondary containment, and to immediately initiate action to restore the Unit 3 SDG to operable status.

Contrary to the above, from February 1, 2023, to January 30, 2024, with each unit in either Mode 4 or 5 at least once or with movement of recently irradiated fuel in the secondary containment, one Unit 3 SDG was inoperable and the licensee failed to immediately suspend core alterations, immediately suspend movement of recently irradiated fuel, and to immediately initiate action to restore the Unit 3 SDG to operable status.

Browns Ferry Nuclear Plant, Units 1, 2, and 3 TS Section 3.0, 'LCO Applicability', Subsection 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, between February 1, 2023, to January 30, 2024, Units 1, 2, and 3 entered a TS 3.8.1 applicable mode when LCO TS 3.8.1 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.

Licensee-Identified Non-Cited Violation 71153 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: 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 March 16, 2014, until February 21, 2024, the licensee failed to establish design control measures to verify or check the adequacy of design of the Unit 3 inboard main steam drain line isolation valve, 3-FCV-001-0055. Specifically, the licensee failed to ensure that the motor operated valve limit and torque switch settings associated with 3-FCV-001-0055 would ensure that the valve would perform its intended safety function under all design basis conditions.

Browns Ferry Nuclear Plant, Unit 3 TS Subsection 3.6.1.3, 'Primary Containment Isolation Valves (PCIVs),' Condition A, requires, in part, that requires that with one or more penetration flow paths with one PCIV inoperable, Action A.1 - Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. TS 3.6.1.3 Condition E requires that with the required action and associated Completion time of Condition A not met, place the unit 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, one required PCIV was inoperable from March 16, 2014, to February 16, 2024, 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 Plant, Unit 1 TS Section 3.0, 'LCO Applicability', Subsection 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, between March 16, 2014, and December 10, 2022, BFN Unit 3 entered a TS 3.6.1 applicable mode on multiple occasions when LCO TS 3.6.1 required actions were not met.

Significance/Severity: Green. Severity Level IV. The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The affected cornerstone was barrier integrity, as determined by IMC 0609, Attachment 4, Initial Characterization of Findings. The inspectors screened the performance deficiency using Exhibit 3 of Appendix A. The inspectors answered no to both questions C1 & C2, therefore the issue screened Green. This violation is characterized as a Severity Level IV NCV based on its similarity to SLIV 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 outboard main steam drain line isolation valve, 3-FCV-001-0056, remained capable of performing its required safety function for all applicable periods of time while 3-FCV-001-0055 was not.

Corrective Action References: CR

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On July 9, 2024, the inspectors presented the integrated inspection results to E. Quinn Leonard, Plant Manager and other members of the licensee staff.

• On April 11, 2024, the inspectors presented the Emergency Preparedness baseline program inspection exit meeting inspection results to D. Komm and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Work Orders WO 123543722

71111.07A Work Orders 123420296

71111.15 Corrective Action CRs 1902390,

Documents 1920472,

26099,

1445711,

1940354,

1917980,

1938714,

1940394

Drawings Drawing 3- Wiring Diagram 4160V Shutdown Aux Power Schematic

45E766-16 Diagram

Drawing 3- Wiring Diagram Diesel Generators Schematic Diagram

45E767-5

Miscellaneous Purchase orders

24454,

288622

Work Orders WO 121604907

71111.18 Corrective Action CR 1911827

Documents

Drawings Drawing 3- Wiring Diagram 480V Shutdown Aux Power Schematic Rev. 15

45E779-12 Diagram

71111.24 Work Orders WO 122901749,

23287037,

23511382,

23405588,

23543722,

23585976,

23511382,

24446965,

23785860,

117608911

Inspection Type Designation Description or Title Revision or

Procedure Date

71114.02 Procedures EPDP-10 Facilitation of the Alert and Notification System and Rev. 11

Notification Tests

EPDP-14 Evaluation of Changes to Alert and Notification Systems Rev. 37

71114.03 Miscellaneous ERO staff training and qualification records

Procedures CECC EPIP-6 CECC Plant Assessment Staff Procedure for Alert, Site Area Rev. 52

Emergency, And General Emergency

EPIP-6 Activation and Operation of the Technical Support Center Rev. 48

(TSC)

EPIP-7 Activation and Operation of the Operations support Center Rev. 42

(OSC)

EPIP-7 Activation and Operation of the Operations Operations Rev. 42

Support Center (OSC)

EPIP-8 Personnel Accountability and Evacuation Rev. 37

TRN-30 Radiological Emergency Preparedness Training Rev. 43

71114.04 Miscellaneous BFN 2023-016 Screening Evaluation Form for EPIP-5, Revision 65 11/02/2023

Procedures CECC EPIP-1 CENTRAL EMERGENCY CONTROL CENTER (CECC) Rev. 78

OPERATIONS

EPDP-17 10 CFR 50.54(q) Evaluations of Emergency Plan Changes Rev. 10

EPIP-1 Emergency Classification Procedure Rev. 68

EPIP-2 Notification of Unusual Event Rev. 46

EPIP-3 Alert Rev. 50

EPIP-4 Site Area Emergency Rev. 50

EPIP-5 General Emergency Rev. 65

REP-Appendix A Browns Ferry Nuclear Plant Radiological Emergency Plan Rev. 113

REP-Generic Radiological Emergency Plan (Generic Part) Rev. 116

71114.05 Corrective Action 1868017 REP Protocol enhancement opportunity 08/06/2023

Documents 1872405 Training Drill-Improvement Opportunities 08/27/2023

1873784 Potential trend related to Emergency Preparedness 09/03/2023

performance

1880245 Missing BFN Joint Information Center Laptop 10/08/2023

1885218 Evaluate TSC Operations Communicator Requirements 10/30/2023

1896853 Training Drill Enhancement Opportunities 12/31/2023

Corrective Action CR 1923873 NRC Inspection BFN Emergency Preparedness

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents OS-0239 R0 Guidance for alternate monitoring when WRGERMS is non 04/11/2024

Resulting from functional

Inspection

Engineering EPDP-1 Procedures, Technical Reports, Maps, and Drawings Rev. 22

Evaluations KLD TR-1358 2023 Evacuation Time Estimates Analysis Rev. 0

Miscellaneous Mutual Aid Letters of Agreement (LOAs) with Offsite

Response Organizations

Browns Ferry Wide Range Gaseous Effluent Radiation

Monitor Summary Package - 0-RM-090-0306

Signed Drill and Exercise Program critiques

Attach 2 TEENS Test Performance Documents and Records

Procedures 0-TI-67 Determination of Stack Release Rates Rev. 9

EPDP-3 Emergency Plan Exercises and Preparedness Drills Rev. 24

EPIP-16 Termination and Recovery Procedure Rev. 9

EPIP-8 Personnel Accountability and Evacuation Rev. 37

NPG-SPP-18.3.5 Equipment Important to Emergency Response Rev. 0014

NPG-SPP-22.102 Self-Assessment and Benchmarking Programs Rev. 14

NPG-SPP-22.300 Corrective Action Program Rev. 0024

NPG-SPP-22.600 Issue Resolution Rev. 0013

Self-Assessments SSA 2301 Audit of FC Emergency Preparedness (EP) 01/24/2023

71151 Miscellaneous 2023 Q2-4 Data Drill and Exercise Performance (DEP), Emergency

Response Organization (ERO), and Alert Notification (ANS)

Performance Indicator Records

Procedures EPDP-11 Emergency Preparedness Performance Indicators Rev. 12

71153 Corrective Action CR 1911827

Documents

Miscellaneous LER

05000296/2024-

001-00

Work Orders WO 124337207,

20588210,

117054961,

21335280

19