IR 05000259/2022001

Integrated Inspection Report 05000259/2022001, 05000260/2022001, 05000296/2022001 and Exercise of Enforcement Discretion

ML22131A086
Person / Time
Site:	Browns Ferry
Issue date:	05/12/2022
From:	Tom Stephen NRC/RGN-II/DRP/RPB5
To:	Jim Barstow Tennessee Valley Authority
References
EA-22-023 IR 2022001
Download: ML22131A086 (27)

Text

May 11, 2022

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2022001, 05000260/2022001, 05000296/2022001 AND EXERCISE OF ENFORCEMENT DISCRETION

Dear Mr. Barstow:

On March 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On April 21, 2022, the NRC inspectors discussed the results of this inspection with Mr. Joe Quinn and other members of your staff. The results of this inspection are documented in the enclosed report.

One Severity Level IV violation without an associated finding 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.

No NRC-identified or self-revealing findings were identified during this inspection.

If you contest the violation or the significance or severity of the violation 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.

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, Thomas A. Stephen, Chief 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: 05000259/2022001, 05000260/2022001 and 05000296/2022001

Licensee: Tennessee Valley Authority

Facility: Browns Ferry Nuclear Plant

Location: Athens, Alabama

Inspection Dates: January 01, 2022 to March 31, 2022

Inspectors: S. Downey, Senior Reactor Inspector N. Karlovich, Resident Inspector M. Kirk, Senior Project Engineer A. Nielsen, Senior Health Physicist W. Pursley, Health Physicist J. Steward, Senior Resident Inspector

Approved By: Thomas A. Stephen, Chief 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

Unit 1 and Unit 2 480 Volt Load Shed Logic Inoperable Longer than Allowed by Technical Specifications due to Failed Relay Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259,05000260,05000296/202200 1-01 Open/Closed A self-revealed Severity Level (SL) IV Non-Cited Violation (NCV) of Technical Specification (TS) 3.8.1 Conditions C, I, and J for units 1 and 2 was identified when the licensee discovered on September 23, 2021, while performing surveillance test 0-SR-3.8.1.8(I), 480 V Load Shedding Logic System Function Test (Division I), that the Division I load shed relay 0-RLY-231-00A7 would not have energized when required due to an open circuit on the coil, rendering the Division I load shed logic inoperable.

Additional Tracking Items

Type Issue Number Title Report Section Status EDG EA-22-023 Failure to Comply with 10 71124.08 Closed CFR 37 for the Protection of Disused Steam Dryer Pieces Stored in a Concrete Module LER 05000259,05000260/20 50-259/2021-001-00 480 71153 Closed 21-001-00 Volt Load Shed Logic Inoperable Longer than Allowed by Technical Specifications due to Failed Relay

PLANT STATUS

Unit 1 began the inspection period at 100 percent rated thermal power (RTP). On January 14, 2022, the unit performed an orderly shutdown to commence a planned forced outage (F108)primarily to identify and repair the source of increased unidentified drywell floor drain leakage and also to remove a defective fuel assembly from the reactor core. The Unit entered Mode 2 and commenced a reactor startup on January 23, 2022. The Unit was placed in Mode 1 and the main generator synchronized to the grid on January 24, 2022. The Unit ascended in power and following a pre-conditioning ramp rate and several rod pattern adjustments, the Unit was returned to RTP on January 27, 2022, where it remained through the end of the inspection period.

Unit 2 began the inspection period at RTP. On February 18, 2022, the Unit performed a planned downpower to 70 percent RTP to perform a control rod sequence exchange and other planned testing. Unit 2 returned to RTP on February 20, 2022 and remained at or near RTP through the remainder of the inspection period.

Unit 3 began the inspection period at 93 percent RTP, limited by maximum core flow of 104.5 percent, as the Unit approached coast down in preparation for commencing a planned refueling outage (3R20). On January 7, 2022, Unit 3 performed final feedwater temperature reduction and removed the high pressure feedwater heaters from service to coast down to the start of

3R20. On February 25, 2022, Unit 3 performed an orderly shutdown from 75 percent RTP to

commence 3R20. The Unit was in cold shutdown (Mode 4), nearing the completion of the refueling outage at the end 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.

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather of high winds and a tornado watch on February 17, 2022.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

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

(1) Unit 3, Systems used for level control of the reactor cavity and spent fuel pool with fuel pool gates removed on March 8, 2022

(2) Unit 3, Residual heat removal (RHR) loop II following restoration to a shutdown cooling lineup after planned maintenance activities on March 10, 2022

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

(1) The inspectors evaluated system configurations during a complete walkdown of the Alternate Decay Heat Removal System on February 24, 2022.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 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 25-1, which includes elevations 550' and 565' of the intake pump station on January 18, 2022

(2) Fire Area 16, including the Unit 2 Auxiliary Instrument Room, Control Building elevation 593' on February 1, 2022

(3) Fire Area Refuel, Unit 3 Reactor Building elevation 664' on February 24, 2022

(4) Fire Area 16-A, Units 1, 2, and 3 Main Control Rooms on March 28, 2022

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 drill on January 12, 2022

71111.06 - Flood Protection Measures

Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)

The inspectors evaluated internal flooding mitigation protections in the:

(1) Handholes 15 and 26 inspected for submergence of underground cables on January 10, 2022

71111.08G - Inservice Inspection Activities (BWR)

BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding

Activities (IP Section 03.01)

(1) The inspectors evaluated boiling water reactor non-destructive testing by reviewing the following examinations from March 7, 2022 to March 11, 2022:

1. Liquid Penetrant Examination a. RWR-3-019-047, Pipe to Fitting Weld, ASME Class 1. This included a review of the associated welding activities.

b. RWR-3-028-001, Valve to Tubing Connector Weld, ASME Class 1.

This included a review of associated welding activities.

2. Ultrasonic Examination a. DRHR-3-02, Pipe to Valve Weld, ASME Class 2.

b. KR-3-50, Pipe to Elbow Weld, ASME Class 1.

c.

N1B-NV, Reactor Vessel Nozzle to Shell Weld, ASME Class 1.

d. RCRD-3-33, Nozzle to Cap Weld, ASME Class 1.

3. Visual Examination a. N1B-IR, Reactor Vessel Nozzle Inner Radius, ASME Class 1.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the Control Room during the following activities:

Unit 3 Final feedwater temperature reduction (FFTR) during coast down for upcoming refueling outage on January 7, 2022 Unit 1 Shutdown, depressurization and cooldown to planned support forced outage (F108) activities on January 14, 2022 Unit 3 Shutdown, depressurization and cooldown for planned refueling outage on February 25 and February 26, 2022

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

(1) The inspectors observed and evaluated licensed operators' performance in the Unit 3 simulator during scenarios OPL175S294 and OPL175S477 on January 12, 2022

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (1 Sample)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) Loose linkage to STA switch associated with 4kV safety related breaker for the D1 Residual heat removal service water (RHRSW) pump identified on December 17, 2021

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) Elevated site aggregate risk (yellow) from various planned work activities including maintenance work associated with control rod drive pump 1B, which is common to Units 1 and 2 per work orders 119821876 and 121185776, and the Unit 0 Division II 480 Volt Load Shedding Logic System Functional Test Per work order 121396246, which was an infrequently performed test, on January 11, 2022

(2) Unit 1, Review of risk mitigation plans and the outage safety plan for the forced outage covering January 15, 2022 through January 24, 2022

(3) Unit 1 Reactor Pressure Vessel (RPV) Head Lift, High Risk Evolution in support of vessel disassembly activities during F108 on January 16, 2022

(4) Unit 3 review of the risk mitigation plans and the outage safety plan for the scheduled refueling outage (3R20) on February 25, 2022

(5) Unit 3, Heavy lift of the RPV head in a critical lift zone on February 27, 2022

(6) Unit 3, High decay heat loads 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown and NFPA 805 high risk evolution with lowered inventory on February 26, February 27, and February 28, 2022

(7) Unit 3, RHR loop II pump and minimum flow valve logic testing, which was a phoenix Probabilistic Risk Assessment (PRA) risk yellow activity, on February 10, 2022

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) Unit 1, Reactor coolant system following an elevated heat up rate during an ASME pressure test on January 22, 2022

(2) Units 1, 2 and 3, C3 RHRSW pump bearing oil levels lower than described in procedure on January 21, 2022

(3) Unit 1, Past operability evaluation for the reactor coolant pressure boundary documented under condition report 1757875, and completed on February 25, 2022

(4) Unit 3 Scram Insertion Times Potential AC Failure documented under CR 1759461 on March 8, 2022

(5) Unit 3, Alarm received for Diesel Generator 3C battery charger voltage being lower than battery voltage documented in CR 1745380 on January 4, 2022

(6) Unit 3, Control Room Radiation Monitor, BFN-0-RM-090-0259B found deenergized documented under condition report 1765764 on March 30, 2022

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)

The inspectors evaluated the following temporary or permanent modifications:

(1) Unit 1, Temporary modification (TMOD BFN-1-2022-074-002-AWA-001) implemented for a pipe leak on code class 1 piping on January 19, 2022

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (6 Samples)

The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:

(1) Unit 3, 3B Diesel generator (DG) following completion of 2 year and 4 year maintenance activities on January 9, 2022

(2) Unit 1, Core Spray 1A Room Cooler following work to correct high vibration noted during surveillance testing on January 11,2022

(3) Unit 3, Reactor Building Floor Drain Sump Pump Motor 3A following work to correct blown control power fuses on several occasions on January 11, 2022

(4) Unit 1, Main Steam Line "A" Relief Valve BFN-PCV-001-0179 manual cycle test following replacement of the pilot valve on January 24, 2022

(5) Standby Gas Treatment Train A Inlet Damper Actuator Motor replacement and post maintenance test on February 23, 2022

(6) Unit 3, ASME OM Code Testing of Relief Valve, 3-RFV-074-0587D, RHR HTX D Pressure Relief Valve on March 24, 2022

71111.20 - Refueling and Other Outage Activities

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

(1) (Partial)

The inspectors evaluated the Unit 1 Forced Outage (F108) activities which included positively identifying and repairing the source of drywell unidentified leakage and removal of a defective fuel assembly from the reactor core from January 14, 2022 to January 24, 2022.

(2) The inspectors evaluated the Unit 3 Refueling Outage (3R20) planned activities, including installation of new high pressure feedwater heaters 3A1, 3A2, 3C1 and 3C2 from February 26, 2022 to March 31, 2022.

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:

Surveillance Tests (other) (IP Section 03.01) (9 Samples)

(1) Unit 1 and 2, 480V division II load shed logic testing conducted January 11-14, 2022

(2) Unit 1 F108 Core Verification following removal of defective fuel bundle for analysis at a date to be determined on January 21, 2022

(3) Unit 3 Quarterly Check of Shutdown Board 3EB Battery on February 10, 2022

(4) Unit 3, 3B Diesel generator load acceptance testing performed on February 15, 2022

(5) Unit 3, Cooldown rate monitoring of the RPV following unit shutdown on February 25, and February 26, 2022

(6) Unit 1 Standby Liquid Control (SLC) pump functional test performed on February 7, 2022

(7) Unit 3 Reactor Pressure Vessel Hydrostatic Test on March 28, 2022

(8) Unit 3, 3A Diesel generator load acceptance testing performed on February 27 and 28, 2022

(9) Unit 3, Group 6 Primary Containment Isolation System Logic test performed on March 1, 2022

Inservice Testing (IP Section 03.01) (2 Samples)

(1) Unit 1 and 2, Control Bay chilled water pump A and check valves performance test on February 1, 2022

(2) Unit 1 and 2, RHR Service Water Pump A2 quarterly pump test on February 14, 2022

Containment Isolation Valve Testing (IP Section 03.01)

Unit 3, Suppression Chamber Vacuum Relief Valve 3-FCV-64-20 and suppression chamber vacuum breaker 3-CKV-64-800 primary containment local leak rate test on March 10, 2022

71114.06 - Drill Evaluation

Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)

(1) The inspectors reviewed and evaluated the emergency response organization drill response during a Unit 2 shutdown Mode 5 scenario that required the facilities to correctly classify the event and notify the required outside organizations within the required timeframes on February 2,

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards during the licensee's unit 3 refueling outage number 20 (3R20).

Instructions to Workers (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards during 3R20.

Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)

The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:

(1) Licensee surveys of potentially contaminated material leaving the radiological controlled area (RCA) at the main RCA control point during 3R20.

(2) Licensee surveys of potentially contaminated material leaving the RCA at the railroad bay rollup door during 3R20.

(3) Workers exiting the RCA at the main radiation control point during 3R20.

Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)

The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:

(1) Radiation work permit (RWP)-22380031, "3R20 Drywell Carpenter Support," pre-work activities in support of rebuild of plant valve BFN-3-FCV-068-0079 in a High Radiation Area (HRA)"

(2) RWP-22380093, "3R20 Drywell Inspection (ISI/FAC) Activities: Locked High Radiation Area (LHRA}"

(3) RWP-22360052, "3R20 Turbine Building General Maintenance Activities: HRA." Unit 3 Steam Tunnel work.

(4) As Low as Reasonably Achievable Plan 22-0323, 3R20 Rebuild BFN-3-FCV-068-0079

High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)

The inspectors evaluated licensee controls of the following High Radiation Areas.

(1) LHRA Alternate Decay Heat Removal Heat Exchanger Area on the refuel floor

(2) LHRA on a ladder lock on the Unit One Turbine Building Narrow Side Mezzanine

(3) LHRA control at Unit 2 Moisture Separator 2A1, 2B1 and 2C1 West cubicle.

(4) LHRA control at U2 Steam Packing Exhauster cubicle.

(5) LHRA control at U2 Steam Jet Air Ejector cubicle.

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)

(1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements during the licensee refueling outage 3R20.

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &

Transportation

Radioactive Material Storage (IP Section 03.01)

The inspectors evaluated the licensees performance in controlling, labeling and securing the following radioactive materials:

(1) Materials stored at the low level waste onsite storage facility.

(2) Materials stored in the radioactive waste processing building.

Radioactive Waste System Walkdown (IP Section 03.02) (2 Samples)

The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:

(1) Phase separator tanks.

(2) Resin and sludge fill heads.

Waste Characterization and Classification (IP Section 03.03) (3 Samples)

The inspectors evaluated the following characterization and classification of radioactive waste:

(1) 2020 dry active waste.

(2) 2020 reactor water cleanup resin.

(3) 2021 reactor water cleanup resin.

Shipment Preparation (IP Section 03.04) (1 Sample)

(1) The inspectors observed the preparation of radioactive shipment RW 22-049, dewatered resin, low specific activity.

Shipping Records (IP Section 03.05) (4 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) RW 21-014, dewatered resin, Type B

(2) RW 21-205, irradiated metals, Type B

(3) RW 21-080, control rod drives, low specific activity

(4) RW 21-177, irradiated metals, Type B

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===

(1) Unit 1 (January 1, 2021 through December 31, 2021)

(2) Unit 2 (January 1, 2021 through December 31, 2021)

(3) Unit 3 (January 1, 2021 through December 31, 2021)

BI02: RCS Leak Rate Sample (IP Section 02.11) (3 Samples)

(1) Unit 1 (January 1, 2021 through December 31, 2021)

(2) Unit 2 (January 1, 2021 through December 31, 2021)

(3) Unit 3 (January 1, 2021 through December 31, 2021)

OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) April 17, 2021 through January 31, 2022 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1) June 12, 2021 through January 31, 2022

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 50-259 and 260/2021-001-00, 480 Volt Load Shed Logic Inoperable Longer than Allowed by Technical Specifications due to Failed Relay. The circumstances surrounding this LER are documented in the results section.

INSPECTION RESULTS

Enforcement Enforcement Action EA-22-023: Failure to Comply with 10 CFR 71124.08 Discretion 37 for the Protection of Disused Steam Dryer Pieces Stored in a Concrete Module

Description:

On January 28, 2021, the NRC issued Inspection Report numbers 05000259/2020004, 05000260/2020004, and 05000296/2020004, which documented a violation of 10 CFR Part 37.11 involving disused steam dryer pieces (Category 2 material, exempt waste) stored in a large concrete storage module. The violation met the criteria for Enforcement Discretion as described in Enforcement Guidance Memorandum (EGM) 14-001, "Interim Guidance for Dispositioning 10 CFR Part 37 Violations with Respect to Large Components or Robust Structures Containing Category 1 or Category 2 Quantities of Material at Power Reactor Facilities Licensed Under 10 CFR Parts 50 and 52." On March 1, 2022, the inspectors re-evaluated storage and security of the steam dryer pieces and determined that there have been no changes since the last inspection.

Corrective Actions: The licensee has documented the issue in their corrective action program. As specified in EGM 14-001, the application of discretion is authorized until the underlying technical issue is dispositioned through rulemaking or other regulatory action.

Corrective Action References: CR 1646655 and CR 1651677

Enforcement:

Significance/Severity: No safety significance.

Violation: On January 28, 2021, a violation of 10 CFR Part 37.11 was documented in Inspection Report numbers 05000259/2020004, 05000260/2020004, and 05000296/2020004. On March 1, 2022, the inspectors determined that the previously identified violation remains.

Basis for Discretion: This violation continues to meet the criteria for Enforcement Discretion as described in EGM 14-001.

Unit 1 and Unit 2 480 Volt Load Shed Logic Inoperable Longer than Allowed by Technical Specifications due to Failed Relay Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV Applicable 05000259,05000260,05000296/2022001-01 Open/Closed A self-revealed Severity Level (SL) IV Non-Cited Violation (NCV) of Technical Specification (TS) 3.8.1 Conditions C, I, and J for units 1 and 2 was identified when the licensee discovered on September 23, 2021, while performing surveillance test 0-SR-3.8.1.8(I), 480 V Load Shedding Logic System Function Test (Division I), that the Division I load shed relay 0-RLY-231-00A7 would not have energized when required due to an open circuit on the coil, rendering the Division I load shed logic inoperable.

Description:

The Browns Ferry Unit 1 and 2 TS Limiting Conditions For Operation (LCO)3.8.1(b), AC Sources - Operating requires in part that the Unit 1 and 2 diesel generators with two divisions of 480 V load shed logic and common accident signal logic be operable in Modes 1, 2, and 3. On September 23, 2021, the Tennessee Valley Authority identified through surveillance testing that the Division I load shed relay 0-RLY-231-00A7 would not energize when required, which resulted in the inoperability of Division I of the 480 V load shed logic system. The licensee restored the division back to operability on September 25, 2021. The licensee confirmed the last successful test of the relay prior to failure occurred on March 3, 2021, and determined the division was inoperable for a total of 206 days. The 480 V load shed logic divisions are redundant and shared between Units 1 and 2, so that either Division 1 or Division 2 can perform the load shedding required for both units.

Unit 2 was in Mode 5 on March 3, 2021, executing a scheduled refueling outage. Unit 2 entered Mode 2 from Mode 4 on April 21, 2021 and entered Mode 1 on April 22, 2021.

With one division of 480 V load shed logic inoperable for Units 1 and 2, TS 3.8.1 Condition C required action is to restore the required division of 480 V load shed logic to operable status in a completion time of 7 days. With required action and associated completion time of Condition C not met, the Condition I.1 required action is to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and I.2 be in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

TS 3.8.1 Condition J in part requires a unit to enter LCO 3.0.3 immediately whenever two divisions of 480 V load shed logic are inoperable. Unit 1 met the entry conditions for TS 3.8.1 condition J twice during the time frame that the Division I relay was inoperable for a total of 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />, due to testing which made the Division II logic inoperable. Since it was unknown to the licensee at the time that the Division I was inoperable, LCO TS 3.0.3 was not entered thus the TS required action was not met.

Browns Ferry Nuclear Plant, Unit 2 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. However, on April 21, 2021, when Unit 2 entered Mode 2, the unit entered a TS 3.8.1 applicable mode when LCO TS 3.8.1 required actions were not met, and the associated actions of TS 3.8.1 did not permit continued operation in the mode for an unlimited period of time. This failure to meet the requirements of LCO 3.0.4 was identified during the review of the licensee event report by NRC inspectors.

Unit 2 also satisfied entry conditions twice for TS 3.8.1 Condition J during that time frame due to testing which made the Division II logic inoperable, the first time was from March 4 to March 6, 2021, when Unit 2 was in Mode 5. However, TS 3.8.1 is only applicable when in Modes 1, 2, or 3. During the second time frame of May 19 to May 20, for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, Unit 2 was in Mode 1, a TS 3.8.1 LCO applicable Mode.

While entry into TS 3.0.3 and Mode 4 was not met for both units, this equipment configuration was of low risk to both Units. As documented in LER 50-259/2021-001-00, the licensee performed a Probabilistic Risk Analysis (PRA) for both divisions of 480 V load shed logic inoperable for 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />. Upon NRC request, the licensee also performed a PRA for inoperability of the Division I relay for 206 days. Additionally, the NRC performed their own risk analysis. In all cases the risk was determined as very low significance.

Corrective Actions: The licensee replaced the Division I relay and performed a Part 21 analysis and evaluation. The corrective actions determined that Part 21 was not applicable for the failed relay.

Corrective Action References: Condition Report (CR) 1723229

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: The inspectors used the guidance in Inspection Manual Chapter 0612 Appendix B, Issue Screening Directions dated October 1, 2021, to screen the violation. The inspectors determined the there was no potential willfulness. Following the Reactor Oversight Process (ROP) path in Figure 1, Issue Screening in Appendix B, the inspectors determined that there was no performance deficiency and therefore no finding because the issue was not reasonably foreseeable and preventable by the licensee.

Following the traditional enforcement path in Figure 1, Appendix B, the inspectors determined that traditional enforcement applied. The violation was assessed in accordance with the NRC Enforcement Policy, dated January 14, 2022. Section 2.2.1 of the enforcement policy states in part that Whenever possible, the NRC uses risk information in assessing the safety or security significance of violations and assigning severity levels. A higher severity level may be warranted for violations that have greater risk, safety, or security significance, while a lower severity level may be appropriate for issues that have lower risk, safety, or security significance.

Since a violation of technical specifications was identified for the 480V load shed system, risk analyses were performed by a regional Senior Reactor Analyst (SRA) for two scenarios. The first scenario was for the 206 days loss of Division I load shed logic and, the second scenario was for the 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> when both Division I and Division 2 were out of service. These analyses were performed using Systems Analysis Programs for Hands-on Integrated Reliability Evaluation (SAPHIRE) 8 version 8.2.5 and the Browns Ferry Unit 1 Standardized Plant Analysis Risk (SPAR) model version 8.61 dated June 11, 2019. For the first scenario of the loss of Division I load shed logic for 206 days, the SRA conservatively assumed that there was a risk associated with a common cause failure of the Division 2 load shed relay. In the second scenario for the 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> when Division I and II were both out of service, the SRA conservatively assumed that the load sequencer on all four diesels would fail to operate due to the common cause failure of both relays being inoperable. The dominant accident sequence in both cases was a loss of offsite power due to weather related reasons. For unit 2, the inspectors considered the unit 1 analysis to be bounding. The change on core damage frequency was 1.56E-7 events or less per year which corresponds to a finding of very low safety significance (Green) if evaluated under the ROP. Additionally, the licensees PRA analysis also indicated the risk in both scenarios as very low. Therefore, the inspectors characterized this issue as a SL IV NCV, since no performance deficiency was identified.

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

Operating,' Condition C, requires that with one division of 480 V load shed logic inoperable the division must be restored to operable status within 7 days. Condition I requires in part that if the required action and associated completion time of Condition C is not met then the unit must 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 be in 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 conditions C and I, for Unit 1 the required division was inoperable from March 3, 2021, to September 25, 2021, and the Unit did not enter Mode 3 in 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 for Unit 2, the required division was inoperable from March 3, 2021, to September 25, 2021, and once the TS was applicable when Unit 2 entered Mode 2 on April 21, 2021, 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 Units 1 and 2 TS Section 3.8.1 Condition J, requires in part immediate entry into TS LCO 3.0.3 when two divisions of 480 V logic are declared inoperable while in Modes 1, 2, or 3. Contrary to the above, for Unit 1 on March 4 to March 6, 2021, and May 19 to May 20, 2021, both divisions were inoperable for approximately 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />, and LCO TS 3.0.3 was not entered. Contrary to the above for Unit 2, on May 19 to May 20, 2021, both divisions were inoperable approximately 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, and LCO TS 3.0.3 was not entered.

Browns Ferry Nuclear Plant, Unit 2 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, on April 21, 2021, Unit 2 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.

Minor Violation 71153 Minor Violation: During inspector review of LER 50-259/2021-001-00, a minor of violation of 10 CFR 50.73(a)(2)(i)(B) was noted. 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the licensee shall report any operation or condition which was prohibited by the plants Technical Specifications (TS). NPG-SPP-03.5, Regulatory Reporting Requirements, Section 3.2.3.A.1 requires in part that the organization or individual responsible for preparing the report shall ensure that the reportaddresses all necessary items.

Contrary to the above, on November 22, 2021, the licensee submitted a License Event Report that did not include a condition prohibited by Unit 2s TS. Specifically, the LER did not identify that Unit 2 did not meet TS LCO 3.0.4 when it changed from Mode 4 to Mode 2. Unit 2 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. The inspectors do not consider the failure to report TS 3.0.4 applicability to be administrative in nature.

Screening: The inspectors determined the performance deficiency was minor. The inspectors determined the violation to be minor because, in accordance with Inspection Manual Chapter (IMC) 0612 Appendix B, Additional Issue Screening Guidance, traditional enforcement does not apply, and the performance deficiency (PD) does not meet any of the More-than-Minor (MTM) criteria. The PD did not meet any of the MTM criteria because, in this instance, the failure to meet TS 3.0.4 did not increase the significance of the TS violations described in LER 50-259/2021-001-00.

Enforcement:

The licensee has entered the issue into their corrective action program CR 1770904, to restore compliance. This failure to comply with 10 CFR 50.73(a)(2)(i)(B)constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

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

On April 21, 2022, the inspectors presented the integrated inspection results to Mr. Joe Quinn and other members of the licensee staff.

On March 4, 2022, the inspectors presented the Baseline Radiation Protection Exit Meeting inspection results to M. Rasmussen and other members of the licensee staff.

On March 10, 2022, the inspectors presented the Baseline ISI Exit Meeting inspection results to Mr. Daniel Komm, Assistant Plant Manager and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Procedures 0-AOI-100-7 Severe Weather Revision 47

NPG-SPP-07.1.8 Severe Weather and Natural Disasters Revision 2

71111.04 Drawings 0-15E740-1 Single Line Diagram ADHR Service Entrance and MCC Rev 13

0-47E830-2 Flow Diagram Radwaste Rev 29

0-47E832-1 Flow Diagram Fuel Pool Filter/Demineralizer System Rev 14

0-47E873-1 Flow Diagram Auxiliary Decay Heat Removal System Rev 8

Sheet 1

0-47E873-2 Flow Diagram Auxiliary Decay Heat Removal System Rev 8

Sheet 2

1-47E818-1 Flow Diagram Condensate Storage and Supply System Rev 38

3-47E811-1 Flow Diagram Residual Heat Removal System Rev 73

3-47E814-1 Flow Diagram Core Spray System Rev 35

3-47E818-1 Flow Diagram Condensate Storage and Supply System Rev 29

3-47E832-1 Flow Diagram Fuel Pool Filter/Demineralizer System Rev 3

3-47E855-1 Flow Diagram Fuel Pool Cooling Equipment Rev 26

3-47E858-1 Flow Diagram RHR Service Water System Rev 37

Miscellaneous BFN-22 Section 10.22 Final Safety Analysis Report for the Auxiliary Decay 02/24/2022

Heat Removal System (ADHR)

Procedures 0-OI-72/ATT-1 Auxiliary Decay Heat Removal System Valve Lineup Rev 55

Checklist

0-OI-72/ATT-2 Auxiliary Decay Heat Removal System Panel Lineup Rev 54

Checklist

0-OI-72/ATT-3 Auxiliary Decay Heat Removal System Electrical Lineup Rev 55

Checklist

3-OI-74 Residual Heat Removal System Rev 135

3-OI-75 Core Spray System Rev 75

3-POI-78 Reactor Water Letdown During Refueling Outages Rev 11

3-POI-78-1 Compensatory Decay Heat Removal During Refuel Rev 0

Operation using Condensate Transfer Pumps through

the Core Spray Injection Line

71111.05 Fire Plans FPR Volume 2 Fire Protection Report Volume 2 Rev 70

Inspection Type Designation Description or Title Revision or

Procedure Date

Miscellaneous OPDP-3-1 Fire Drill Evaluation Report 09/08/2021

Procedures 0-AOI-26-1 Fire Response Rev 23

FP-0-000-INS001(D) Inspection of Portable and Wheel Type Fire Extinguisher Rev 29

Stations (Protected Area Ancillary Bldgs)

NPG-SPP-18.4.7 Control of Transient Combustibles Rev 15

OPDP-3 Administration of Pre-fire Plans, Fire Emergency Rev 0

Response, and Development and Evaluation of Fire

Drills

YD-DGB Attachment 22 Fire Protection Report Volume 2 Rev 69

71111.06 Work Orders 121856934 Sump Pump Operability Check Handholes 15 and 26 01/10/2022

71111.08G NDE Reports R-040 Ultrasonic Examination of Weld KR-3-50 03/07/2022

R-041 Ultrasonic Examination of N1B-NV 03/08/2022

R-071 Ultrasonic Examination of Weld No. RCRD-3-33 03/18/2022

R-072 Ultrasonic Examination of Weld No. DRHR-3-02 03/12/2022

Work Orders Work Orders (by 119943742, 120511844

number)

71111.11Q Miscellaneous Simulator Exercise Guide for OPL175S294 Rev 2

Simulator Exercise Guide for OPL175S477 Rev 0

Procedures 1-AOI-100-1 Reactor Scram Rev 30

1-GOI-100-12 Power Maneuvering Rev 19

1-GOI-100-12A Unit Shutdown from Power Operation to Cold Shutdown Rev 36

and Reductions in Power During Power Operations

1-OI-47 Turbine Generator System Rev 71

3-AOI-85-5 Rod Drift In Rev 15

3-EOI-1 RPV Control - Modes 1-3 Unit 3 Browns Ferry Nuclear Rev 15

Plant

3-EOI-2 Primary Containment Control Unit 3 Browns Ferry Rev 15

Nuclear Plant

3-OI-6 Feedwater Heating and Misc Drains System Rev 99

EPIP-1 Attachment 1 BFN Hot Condition ICs/EALs Rev 61

NPG-SPP-17.8.4 Conduct of Simulator Operations Rev 9

OPDP-1 Conduct of Operations Rev 53

71111.12 Corrective Action CR 1742947,1748454,

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents 1744459, 1389131,

144272, 1654505

Drawings 0-45E766-23 Wiring Diagram 4160V Shutdown Aux Power Schematic Rev 58

Diagram

3-45E766-9 Wiring Diagram 4160V Shutdown Auxiliary Power Rev 41

Schematic

Miscellaneous December Maintenance Rule Report 01/28/2022

BFN-VTD-SI06-0040 Installation, Operation and Maintenance Instructions for Rev 5

Siemens 5KV Horizontal Vacuum Circuit Breaker

Procedures EPI-0-000-BKR015 4KV Wyle/Siemens Horizontal Vacuum Circuit Breaker Rev 46

(Type-3AF) and Compartment Maintenance

NPG-SPP-06.2 Preventive Maintenance Rev 14

NPG-SPP-09.26.15 Medium and Low Voltage Circuit Breaker Testing and Rev 1

Maintenance Program

Work Orders WO 122599426,

20681854

71111.13 Corrective Action CR 1747621, 1757737,

Documents 1758938

Engineering EWR 18CEB111259 Provide Clarification for the Design and Licensing Basis 12/12/2018

Evaluations for the Operation of the Reactor Building Overhead

Crane at Temperatures below 60.5 F

Miscellaneous Unit 1 Midcycle F108 Outage Safety Plan Rev 0

3R20 Outage Safety Plan Rev 0

Procedures 0-SR-3.8.1.8(II) 480V Load Shedding Logic System Functional Test Rev 19

(Division II)

0-TPP-FPP-011 BFN NFPA 805 Non-Power Operations Fire Risk Rev 6

Management

3-SR-3.3.5.1.6(B II) Functional Testing of RHR Loop II Pump and Minimum Revision 23

Flow Valve Logic

MSI-0-000-LFT001 Lifting Instructions for the Control of Heavy Loads Rev 79

NPG-SPP-07.2.11 Shutdown Risk Management Rev 15

NPG-SPP-07.3 Work Activity Risk Management Process Rev 36

NPG-SPP-10.6 Infrequently Preformed Test or Evolutions Rev 2

TVA-TSP-18.721A TVA Rigging Manual Rev 5

Inspection Type Designation Description or Title Revision or

Procedure Date

TVA-TSP-18.802 Requirements for the Safe Operations Cranes Rev 19

Work Orders WO 121906415 02/10/2022

71111.15 Corrective Action 1759461 and 1758241 03/08/2022

Documents CR 1749275

Miscellaneous RPV and RCS Temperature Data January 22,

22

PORC Item Summary for ASME Section XI System January 18,

Leakage Test of RPV Vessel and associated piping 2022

Past Operability Evaluation Documentation for CR 02/17/2022

1747875

BFN-VTD-G082-0016 Browns Ferry 1, 2, 3 (GEK-113773A) Operation and Rev 1

Maintenance Manual Replacement Residual Heat

Removal Service Water Pump Motor

Procedures 0-GOI-300-1/ATT-12 Outside Operator Round Log Rev 268

1-SI-3.3.1.A ASME Section XI System Leakage Test of the Reactor Rev 25

Pressure Vessel and Associated Piping (ASME Section

III, Class 1 and 2)

1-SR-3.4.9.1(1) Reactor Heatup and Cooldown Rate Monitoring Rev 14

1-SR-3.4.9.1(2) Reactor Vessel Shell Temperature and Reactor Coolant Rev 14

Pressure Monitoring during In-Service Hydrostatic or

Leak Testing

3-SR-3.1.4.1 Scram Insertion Times Rev 36

NPG-SPP-22.207 Procedure Use and Adherence Rev 10

OPDP-8 Operability Determination Process and Limiting Rev 28

Conditions for Operation Tracking

Work Orders 122833321 03/29/2022

71111.18 Drawings 1-47E811-1-ISI ASME Section XI Residual Heat Removal System Code Rev 19

Class Boundaries

Engineering BFN-1-2022-074-002 Cut and Cap Test Line for RHR Valve 1-SHV-74-55

Changes

Miscellaneous Troubleshooting Plan for BFN-1-SHV-074-0794A 01/16/2022

Procedures NISP-IP-ENG-001 Standard Design Process (EB-17-06) Rev 2

NPG-SPP-09.3 Plant Modifications and Engineering Change Control Rev 37

NPG-SPP-09.5 Temporary Modifications Temporary Configuration Rev 19

Inspection Type Designation Description or Title Revision or

Procedure Date

Changes

Work Orders WO 122645957 TMOD BFN-1-2022-074-002-AWA-001 01/16/2022

71111.19 Corrective Action 1746118 3A RBFD Sump Pump blown control power fuse 01/06/2022

Documents 1746301 01/07/2022

CR 1749652 Tools left in Drywell 01/24/20222

Miscellaneous 0-MEG-TRBSHT-001 Initial Troubleshooting Rev 3

0-TI-20 Control Rod Drive System Testing and Troubleshooting Rev 26

Form NPG-SPP-09.2-26 Environmental Qualification Maintenance Work Record 06/30/2014

TVA 40569 Nameplate or Location Data Update Request 11/26/2014

TVA 40630 BFN Replacement EQ Whole Device Field Verification 02/08/2019

Form

TVA 40638 EQ Inspection For Degradation 06/30/2014

TVA 40695 Configuration Control Log for Wire Lifts 08/09/2019

TVA 40928 Foreign Material Control Requirements and Pre-job 12/02/2020

Briefing

Procedures 0-TI-577 Inservice Testing of Pressure Relief Devices Rev 9

0-TI-577(TEST) Inservice Testing of ASME and Augmented Pressure Rev 9

Relief Devices

1-SR-3.4.3.2 Main Steam Relief Valves Manual Cycle Test Rev 10

3-SR-3.8.1.1(3B) Diesel Generator 3B Monthly Operability Test Rev 62

3-SR-3.8.4.2(DG 3B) Diesel Generator 3B Battery Service Test Rev 27

ECI-0-000-CND004 Disconnect and Reconnect of Flexible Conduit Rev 0

ECI-0-000-RAY001 Raychem Removal, Selection and Installation - 600V Rev 1

and Less

EPI-3-082-DGZ03B Diesel Generator 3B 2 Year Inspection Rev 12

MCI-0-000-RLV001 Generic Maintenance Instructions for Relief Valves Rev 70

MPI-0-082-INS003 Standby Diesel Engine 48 Month Inspection Rev 82

Work Orders 122589175 BFN-1-FCV-085-40A/38-27 02/25/2022

2630723 01/07/2022

2647338, 122645986

2721253 02/09/2022

2802583 03/24/2022

71111.20 Corrective Action 1770919 NRC Observation Pre-exit Meeting for Integrated 04/20/2022

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents Inspection Period 2022001 regarding the use of Drones

Resulting from during F108 and 3R20

Inspection

Miscellaneous 1-SR-3.4.9.1(1) BFN Unit 1 Reactor Heatup and Cooldown Rate Rev 14

Monitoring

71111.22 Corrective Action CR 1747074, 1747401,

Documents 1747611, 1748064,

1747533, 1541042,

1714276, 1755096

CR 1757711, 1757714,

1758234

Drawings 0-47E866-9 Flow Diagram Chilled Water Circulating Pumps Rev 14

Engineering 06-1-IST-063-269 Evaluation Form for ASME Section XI IST Test Results 08/28/2006

Evaluations for 1-PMP-063-0006A and 1-PMP-063-0006B

17-0-IST-023-660 Evaluation form for AMSE OM Code IST results for 0- 07/30/2017

PMP-023-005

Miscellaneous BFN-1-229 FC-1 Fuel Assembly Transfer Form 01/21/2022

BFN-1-229 FC-1 Fuel Assembly Transfer Form 01/21/2022

BFN-1-230 Fuel Assembly Transfer Form 01/21/2022

Design Criteria Control Bay and Reactor Building Board Rooms Rev 16

Document, BFN-50- Environmental Control Systems

7030A

Procedures 0-GOI-100-3A Refuel Operations (In-Vessel Operations) 94

0-GOI-100-3B Operations in Spent Fuel Pool Only 69

0-GOI-100-3C Fuel Movement Operations During Refueling 101

0-SI-3.2.33(CW A 1 & 2 Control Bay CHW Pump A Comprehensive Pmp Rev 6

COMP) and Check Vlv Performance Test

0-SI-4.5.C.1(A2) RHRSW Pump A2 IST Group A Quarterly Pump Test Revision 12

0-SR-3.8.1.8(II) 480V Load Shedding Logic System Functional Test Rev 19

(Division II)

0-TI-298 Diesel Generator Operating Data Acquisition Rev 24

0-TI-362 Inservice Testing Program Revision 60

0-TI-362(Bases) IST Program Bases Document Revision 20

0-TI-444 Augmented Inservice Testing Program Rev 15

Inspection Type Designation Description or Title Revision or

Procedure Date

0-TI-444 (BASES) AIST Program Bases Document Rev 9

1-SI-4.4.A.1 Standby Liquid Control Pump Functional Test Revision 34

3-OI-82 Standby Diesel Generator System Rev 154

3-SI-4.7.A.2.g-3/64b Primary Containment Local Leak Rate Test Torus Revision 10

Vacuum Breaker: Penetration X-205

3-SR-3.3.6.2.4 (GRP 6) Group 6 PCIS Logic Revision 19

3-SR-3.4.9.1(1) Reactor Heatup and Cooldown Rate Monitoring Revision 26

3-SR-3.4.9.1(1) Reactor Heatup and Cooldown Rate Monitoring Rev 26

3-SR-3.8.1.9(3A) Diesel Generator 3A Emergency Load Acceptance Test Revision 28

3-SR-3.8.1.9(3B OL) Diesel Generator 3B Emergency Load Acceptance Test Rev 23

With Unit 3 Operating

3-SR-3.8.6.2(3EB) Quarterly Check of Shutdown Board 3EB Battery Rev 17

3.SI-3.3.1.A ASME Section XI System Leakage Test of the Reactor Rev 35

Pressure Vessel and Associated Piping (ASME Section

III, Class 1 and II)

NPG-SPP-05.8 Special Nuclear Material Control 11

Work Orders WO 120674156,

21741051, 119644293

WO 121447445, 03/10/2022

21534136

WO 121471629

WO 121471772

WO 121882175 02/07/2022

WO 122631109 02/14/2022

71124.01 Corrective Action CR 1759208 03/03/2022

Documents

Resulting from

Inspection

Procedures NISP-RP-002 Radiation and Contamination Surveys Rev. 0001

NPG-SPP-05.1.1 Alpha Radiation Monitoring Program Revision 009

Radiation M-20210308-50 68-3 Valve Breach 03/08/2021

Surveys M-20220302-40 Removal of HPCI/RCIC Drain-pipe removal

Survey # M-20220225-MO367 Unit 3 DW 550' Sub-Pile Room 02/25/2022

Inspection Type Designation Description or Title Revision or

Procedure Date

Self-R38 210520 1034 Browns Ferry Nuclear Plant Annual Radionuclide 06/22/2021

Assessments Trending and Assessment Report for 2020

71124.08 Corrective Action CR 1646655

Documents CR 1689399

Miscellaneous Low Level Waste Module Inventory 04/20/2019

71151 Miscellaneous NEI 99-02 Regulatory Assessment Performance Indicator Revision 7

Guideline

PI Summary PI Summary Report for Browns Ferry Units 1, 2, and 3 01/18/2022

for BI01 and BI02

Procedures 1-SR-3.4.6.1 Dose Equivalent Iodine-131 Concentration Revision 8

2-SR-3.4.6.1 Dose Equivalent Iodine-131 Concentration Revision 12

3-SR-3.4.6.1 Dose Equivalent Iodine-131 Concentration Revision 10

Work Orders WO 121762900,

21780742, 121780743,

21741101, 121762899,

21637640, 121637641,

21682603, 121682604,

21168658, 121168660,

21168656, 121168659

71153 Calculations NDN0009992012000013 TVA Fire PRA-Task 7.5 Fire-Induced Risk Model Revision 9

Corrective Action 1723229 06/30/2021

Documents

Corrective Action 1770904 NRC Review of LER 259/2021-001-00 04/20/2022

Documents

Resulting from

Inspection

Drawings 0-45E771-2 Wiring Diagram 480V Diesel Aux Power Schematic Revision 36

Diagram

0-731E734 480 Aux Rly Pnls 25-43A-1, -2 05/30/2007

0-731E734 480V Logic Aux Relay Pnl Revision 5

0-731E750 480V Logic Aux Rly Panels 04/07/2004

0-731E750 480V Logic Aux Rly Panel 24-42A 11/12/2004

0-731E750 480V Logic Aux Rly Panels 08/12/2005

Inspection Type Designation Description or Title Revision or

Procedure Date

Engineering BFN-0-21-132 R1 PRA Evaluation Response for CR 1723229 10/20/2021

Evaluations BFN-0-22-023 R0 Past Operability Evaluation Documentation for CR 03/25/2022

23229

Procedures NEDP-26 Probabilistic Risk Assessment Revision 13

NPG-SPP-03.5 Regulatory Reporting Requirements Revision 17

NPG-SPP-09.11.2 Risk Assessment Methods for Technical Specifications Revision 4

Work Orders 04-715230-000 04/05/2005

24