IR 05000259/2018001

NRC Integrated Inspection Report 05000259/2018001, 05000260/2018001 and 05000296/2018001

ML18128A153
Person / Time
Site:	Browns Ferry
Issue date:	05/08/2018
From:	Masters A Division Reactor Projects II
To:	James Shea Tennessee Valley Authority
References
IR 2018001
Download: ML18128A153 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION May 8, 2018

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000259/2018001, 05000260/2018001, AND 05000296/2018001

Dear Mr. Shea:

On March 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Plant, Units 1, 2, and 3. On April 20, 2018, the NRC inspectors discussed the results of this inspection with Mr. W. Paulhardt and other members of your staff.

The results of this inspection are documented in the enclosed report.

NRC inspectors documented four findings which were determined to be of very low safety significance (Green) in this report. All of these findings involved violations of NRC requirements. Because of their very low safety significance, the NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest any of the violations or significance of these NCVs, 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, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; 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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Anthony D. Masters, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket Nos.: 50-259, 50-260, 50-296 License Nos.: DPR-33, DPR-52, DPR-68

Enclosure:

NRC IIR 05000259/2018001, 05000260/2018001 and 05000296/2018001

REGION II==

Docket Nos.: 50-259, 50-260, and 50-296 License Nos.: DPR-33, DPR-52, and DPR-68 Report No.: 05000259/2018001, 05000260/2018001, and 05000296/2018001 Enterprise Identifier: I-2018-001-0052 Licensee: Tennessee Valley Authority (TVA)

Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3 Location: Corner of Shaw and Nuclear Plant Road Athens, AL 35611 Dates: January 1, 2018 through March 31, 2018 Inspectors: D. Dumbacher, Senior Resident Inspector M. Kirk, Resident Inspector A. Ruh, Resident Inspector A. Nielsen, Senior Health Physicist R. Kellner, Senior Health Physicist R. Carrion, Senior Reactor Inspector S. Monarque, Project Engineer J. Seat, Project Engineer P. Heher, Project Engineer R. Williams, Senior Reactor Inspector G. Crespo, Senior Construction Inspector Approved by: A. Masters, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance by conducting quarterly integrated baseline inspections at Browns Ferry Nuclear Plant, Units 1, 2, and 3 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. NRC and self-revealed findings, violations, and additional items are summarized in the table below.

List of Findings and Violations Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green [H.1] - 71111.19 Systems Non-cited Violation (NCV) 05000259, Resources 260, 296/2018001-01 Closed A self-revealing, Green, NCV of 10 CFR Part 50 Appendix B, Criterion V, was identified when the licensee failed to perform an adequate post-maintenance test in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the 3C diesel generator output breaker did not ensure that all contacts on replacement stationary switches successfully changed state after installation.

Unauthorized Entry into a High Radiation Area (HRA)

Cornerstone Significance Cross-cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000259, 260, 296/2018001-02 Procedure Closed Adherence A self-revealing, Green, NCV of Technical Specification (TS) 5.7.1, was identified for a worker who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3 A & C Residual Heat Removal Heat Exchanger Room using an incorrect Radiation Work Permit and without being briefed on the radiological conditions.

Failure to Implement Controls for Locked High Radiation Area (LHRA) Access Cornerstone Significance Cross-cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV 05000259/260/296/ 2018001-03 Teamwork Opened/Closed A self-revealing, Green, NCV of TS 5.7.2, was identified for the failure to control access to a LHRA. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without Radiological Personnel (RP) present. In doing so, the worker accessed an area with dose rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry.

Inadequate Configuration Control of High Pressure Coolant Injection (HPCI) Valve Design Issues Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green None 71152 - Annual Systems NCV 05000296/2018001-04 Follow-up of Closed Selected Issues A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, was identified when the licensee failed to ensure adequate control of valve design configurations in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control Revision 6. Specifically, the licensee changed, over time, HPCI discharge valve yoke nut and bearing components contrary to original design without documenting or evaluating the changes Additional Tracking Items Type Issue Number Title Report Status Section URI 05000260,296/ Potential Inadequate Weak Link Analysis 71152 Closed 2017008-01 for Unit 2 and Unit 3, HPCI Discharge Valves URI 05000260,296/ Potential Inadequate Commercial Grade 71152 Closed 2017008-02 Dedication of Components in Safety Related Valves URI 05000260,296/ Potential Inadequate Configuration Control 71152 Closed 2017008-03 of the Unit 2 and Unit 3 HPCI Discharge Valves URI 05000296/ Potential Inadequate Operator Response 71152 Closed 2017008-04 to Inadvertent HPCI Injection LER 05000260/2017- Inoperable Primary Containment Isolation 71153 Closed 002-00 Valve Resulting in Condition Prohibited by Technical Specifications LER 05000259/2016- Incorrect Tap Settings for 480 Volt 71153 Closed 004-01 Shutdown Transformer Results in Inoperability of Associated 480V Shutdown Boards

PLANT STATUS

===Unit 1 operated at 100% rated thermal power (RTP) except for a reactor scram related to a turbine control valve partial closure transient on March 18, 2018. The unit returned to 100%

RTP on March 24, 2018, and operated at that level for the remainder of the inspection period.

Unit 2 operated at 100% RTP for the duration of the inspection period.

Unit 3 operated at 100% RTP until a reactor scram occurred on January 10, 2018, related to vibration-induced failure of hydraulic piping for the #2 turbine control valve. The unit returned to 100% power on January 15, 2018, and operated at that level until February 17, 2018. There were two unplanned downpowers during the inspection period due to #3 turbine control valve oscillations, one planned downpower for 3C reactor feed pump maintenance. From February 17, 2018, through March 31, 2018, Unit 3 was shutdown for a planned refueling outage U3R18.

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 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

Seasonal Extreme Weather

The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures on January 2, 2018.

71111.04 - Equipment Alignment

Partial Walkdown

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

(1) Unit 3 Residual Heat Removal (RHR) Loop II Shutdown Cooling alignment on February 18, 2018

(2) Unit 3 Alternate Decay Heat Removal (ADHR) on February 23, 2018

(3) 4160V AC Electrical System on March 10, 2018

(4) Unit 3 Main Steam System on March 15 and 19, 2018 Complete Walkdown (1 Sample)===

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit

===3 Emergency High Pressure Makeup (EHPM) system on March 22, 2018.

71111.05AQ - Fire Protection Annual/Quarterly

Quarterly Inspection

The inspectors evaluated fire protection program implementation in the following selected areas:

(1) Compartment 25-1, Units 1,2, and 3, 550' Intake Pumping Station and 565' Component Cooling Water (CCW) Pump Deck on January 26, 2018

(2) Unit 2 RHR Heat Exchanger 2B, 2D, 2A, and 2C Rooms Elevation 565 and 593, and Area 2-4 - South of Q - Unit 2 Elevation 593 on February 21, 2018

(3) Unit 2 Auxiliary Instrument room, Fire Area 16-M on March 14, 2018

(4) Compartment 26-A, Units 1, 2 and 3 Turbine Building on March 16, 2018 Annual Inspection (1 Sample)===

(1) The inspectors evaluated fire brigade performance on March 6, 2018. The Browns Ferry

===Fire brigade responded to report of smoke coming from a motor for the Unit 1/2 Diesel Building CO2 tank compressor.

71111.06 - Flood Protection Measures

Internal Flooding

(1) The inspectors evaluated internal flooding mitigation protections in the Unit 2 480V Shutdown Board Rooms on February 2, 2018 Cables (1 Sample)===

The inspectors evaluated cable submergence protection in:

===(1) Hand holes 15 and 26 containing underground cables on January 8, 2018

71111.08 - Inservice Inspection Activities

===

The inspectors evaluated boiling water reactor non-destructive testing by observing or reviewing the following examinations from February 28 to March 1, 2018:

(1) Magnetic Particle Examination (MT)a) MT of Weld HPCI-3-009-003 C1R2, Work Order (WO) 117544712, American Society of Mechanical Engineers (ASME) Class 2. This review involved a pressure boundary weld. (Reviewed)

(2) Liquid Penetrant Examination (PT)a) PT of Weld RWCU-3-001-078 C1R0, WO 117656145 ASME Class 1. This review involved a pressure boundary weld. (Reviewed)

(3) Radiographic Examination (RT)a) RT of Weld HPCI-3-009-003 C1R0, WO 117544712, ASME Class 2. This review involved a pressure boundary weld. (Reviewed)b) RT of Weld HPCI-3-009-003 C1R1, WO 117544712, ASME Class 2. This review involved a pressure boundary weld. (Reviewed)c) RT of Weld HPCI-3-009-003 C1R3, WO 117544712, ASME Class 2. This review involved a pressure boundary weld. (Reviewed)

(4) Ultrasonic Test (UT)a) UT Examination Report R-049, Pipe to Elbow Weld, Component ID: DSRHR-3-04.

ASME Class 1. This review involved a pressure boundary weld. (Reviewed)b) UT Examination Report R-085, Nozzle to Shell Weld, Component ID: N3D-NV.

ASME Class 1. This review involved a pressure boundary weld. (Observed)

(5) Visual Test (VT)a) VT Examination Report R-033, Pipe Support, Component ID: 3-478400-099. ASME Class 1. (Reviewed)b) VT Examination Report R-069, Variable Spring Can, Component ID: RHR-3-H-146.

ASME Class 1. (Reviewed)

71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance

Operator Requalification===

The inspectors observed and evaluated a licensed operator requalification exam session for the Group 0 operating crew on the Unit 3 Simulator involving a stuck open main steam relief valve, inadvertent high pressure coolant injection actuation, unit board trip and Anticipated Transient Without Scram (ATWS) with main steam isolation valves open on January 4, 2018.

Operator Performance (1 Sample)===

The inspectors observed and evaluated startup of the Unit 3 reactor on January 12, 2018,

===Unit 3 turbine control valve manipulations and power maneuvering on January 26, 2018, shutdown of Unit 3 on February 17, 2018, and startup of the Unit 1 reactor on March 21, 2018.

71111.12 - Maintenance Effectiveness

Routine Maintenance Effectiveness

The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:

(1) Unit 3 Turbine Stop and Control Valves. Maintenance Rule (MR) Function 047-B and history of vibrations causing problems.

(2) MR Functions for System 575, 4kV Power Supply and Busses in (a)(1) status

71111.13 - Maintenance Risk Assessments and Emergent Work Control

===

The inspectors evaluated the risk assessments for the following planned and emergent work activities:

(1) Planned risk, associated with inoperable Main Bank Battery 3 and Battery Board 3 on January 2, 2018

(2) Emergent work associated with oscillations of the Unit 3 number 3 Control Valve on January 22, 2018

(3) In-office review of proposed Unit 3 refueling outage risk plan

(4) Shutdown risk associated with Unit 3 on shutdown cooling and reactor water level control at 80 inches on February 17, 2018 (day 1) with Time to Boil at 37 minutes

(5) Reactor pressure vessel head lift on February 18, 2018

(6) Shutdown risk associated with Unit 3 during Operations with Potential for Draining the Reactor Vessel (OPDRV) for replacing B Recirculation Pump seals on February 22, 2018

(7) Shutdown risk associated with Unit 3 during OPDRV for replacing 32 control rod drives on February 27, 2018

(8) Yellow shutdown risk during planned maintenance on Unit 3 Division II 4160V boards on March 10, 2018

(9) Yellow shutdown risk on U-1 with short time to boil with Unit 3 still in a refueling outage on March 20, 2018

71111.15 - Operability Determinations and Functionality Assessments

The inspectors evaluated the following operability determinations and functionality assessments:

(1) Unit 2, HPCI valve 73-44, design opening thrust exceeding the bearing rating and the associated operator work around (OWA) on January 22, 2018

(2) Incorrect RHR system pressure gage used for verification of Technical Specification surveillance test requirements (Condition Report (CR) 1372616, 1373852) on January 5, 2018

(3) Turbine Control Valve Fast Closure channel operability with Unit 3 turbine control valve control circuit fuse and wiring changes (CR 1379519, 1382150) on January 22, 2018

(4) APRM 4 fault and 2-out-of-4 voter number 4 operability (CR 1382124) on January 30, 2018

(5) Past operability evaluation for diesel generator 3C load acceptance test failure (CR

===1389131)

71111.18 - Plant Modifications

The inspectors evaluated the following temporary or permanent modifications:

(1) DCN 69424 Replace Unit 3 Condenser Vacuum Pressure Switches with Pressure Transmitters

71111.19 - Post Maintenance Testing

The inspectors evaluated the following post maintenance tests:

(1) Unit 3 4kV Shutdown Board 3EB loss of power logic system test on March 6, 2018

(2) Testing of Unit 3 overhauled motor operated valve 74-53, RHR Loop I Low Pressure Coolant Injection Valve

(3) Testing of replacement Unit 3 Division I Emergency Core Cooling System (ECCS)

Inverter

(4) Local leak rate test of 3-FCV-73-45 HPCI discharge check valve following installation of softer seat material.

(5) Unit 3 Emergency High Pressure Make-Up Basic Pump Recirculation Testing

(6) Surveillance 3-SR-3.1.7.3 Standby Liquid Control System Enriched Sodium Pentaborate Solution Concentration, Quantity Calculation, and ATWS Equivalency Calculation following implementation of modified boron enrichment for Extended Power Uprate

(7) Test of DCN to install parallel auxiliary contact for 3EC 4kv shutdown board normal feeder breaker 1338

(8) Testing of Unit 3 overhauled motor operated valve 74-73, RHR Loop II Test Outboard Isolation Valve

(9) Testing of Unit 3 overhauled motor operated valve 73-2, HPCI Turbine Steam Supply Inboard Primary Containment Isolation Valve

(10) Testing of replacement STA switch on 3EC diesel generator output breaker

71111.20 - Refueling and Other Outage Activities (Partial Sample)

The inspectors evaluated refueling outage U3R18 activities from February 16, 2018 through March 31, 2018. The inspectors completed inspection procedure sections 03.01.a, b, c, d and e.2.

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Routine===

(1) 3-SR-3.8.1.9, (3B OL) Unit 3 EDG load acceptance test, on February 6, 2018,

(2) 3-SR-3.3.1.1.13 APRM 1-4 calibrations per DCN 68463 Stage 4 associated with the Extended Power Uprate (EPU) modification on March 13, 2018 In-service ===

(1) 0-SI-4.5.C.1(A2-COMP) - Residual Heat Removal Service Water (RHRSW) Pump A2 IST Comprehensive Pump on January 2, 2018

(2) 1-SR-3.5.1.6 (RHR II) - Quarterly RHR System Rated Flow Test Loop II, on February 7, 2018

(3) 3-SR-3.1.7.7, Unit 3 Standby Liquid Control system functional test on March 22, 2018

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment

The inspectors evaluated radiological hazards assessments and controls.

Instructions to Workers (1 Sample)===

The inspectors evaluated worker instructions.

Contamination and Radioactive Material Control (1 Sample)

The inspectors evaluated contamination and radioactive material controls.

Radiological Hazards Control and Work Coverage (1 Sample)

The inspectors evaluated radiological hazards control and work coverage.

High Radiation Area and Very High Radiation Area Controls (1 Sample)

The inspectors evaluated risk-significant high radiation area and very high radiation area

===controls.

Radiation Worker Performance and Radiation Protection Technician Proficiency (1 Sample)===

The inspectors evaluated radiation worker performance and radiation protection technician

===proficiency.

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

and Transportation Radioactive Material Storage

The inspectors evaluated the licensees radioactive material storage.

Radioactive Waste System Walk-down (1 Sample)===

The inspectors evaluated the licensees radioactive waste processing facility during plant

===walkdowns.

Waste Characterization and Classification (1 Sample)===

The inspectors evaluated the licensees radioactive waste characterization and

===classification.

Shipment Preparations (1 Sample)===

The inspectors evaluated the licensees radioactive material shipment preparation

===processes.

Shipment Records (1 Sample)===

The inspectors evaluated the licensees non-excepted package shipment records.

===OTHER ACTIVITIES - BASELINE

71151 - Performance Indicator Verification The Resident Inspectors verified licensee performance indicators submittals listed below for the period from January 1, 2017, through December 31, 2017.

(1) Units 1, 2, and 3 Reactor Coolant System Leakage

(2) Units 1, 2, and 3 Reactor Coolant System Activity The inspectors reviewed licensee PI submittals listed below for the period from April 1, 2017, through February 12, 2018. (1 Sample)===

(1) OR01: Occupational Exposure Control Effectiveness

===71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) Unresolved Item (URI) 05000260, 296/2017008-01, Potential Inadequate Configuration Control of the Unit 2 and Unit 3 HPCI Discharge Valves

(2) URI 05000260, 296/2017008-02, Potential Inadequate Commercial Grade Dedication of Components in Safety Related Valves

(3) URI 05000260, 296/2017008-03, Potential Inadequate Configuration Control of the Unit 2 and Unit 3 HPCI Discharge Valves

(4) URI 05000296/2017008-04, Potential Inadequate Operator Response to Inadvertent HPCI Injection

(5) Problem Identification & Resolution and Regulatory Commitments associated with Unit 3 Extended Power Uprate

71153 - Follow-up of Events and Notices of Enforcement Discretion Events

(1) The inspectors evaluated the plant response and licensees response for a Unit 3 reactor scram on January 10, 2018.

(2) The inspectors responded to a Notice of an Unusual Event after a routine search of a work-related vehicle noted a suspicious object underneath the vehicle. It was later determined the suspicious object was a normal part of the vehicle

(3) The inspectors evaluated the plant response and licensees response for a Unit 1 reactor scram on March 18, 2018.

Licensee Event Reports (2 Samples)===

The inspectors evaluated the following licensee event reports (LER) which can be accessed

===at https://lersearch.inl.gov/LERSearchCriteria.aspx:

(1) LER 05000260/2017-002-00, Inoperable Primary Containment Isolation Valve Resulting in Condition Prohibited by Technical Specifications

(2) LER 05000259/2016-004-01, Incorrect Tap Settings for 480 Volt Shutdown Transformer Results in Inoperability of Associated 480V Shutdown Boards

OTHER ACTIVITIES

- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

71004 - Power Uprate Erosion-Corrosion/Flow-Accelerated-Corrosion Monitoring Programs

Inspectors reviewed the Erosion Corrosion/Flow-Accelerated Corrosion (EC/FAC) program in accordance with the guidance contained in NRC Inspection Procedure 49001, Inspection of Erosion-Corrosion/Flow-Accelerated-Corrosion Monitoring Programs dated 12/11/98.

Summary of Power Uprate Inspection Samples Contained in this Report:

Integrated Plant Operations at the Uprated Power Level (Unit 3) (1 sample)===

(1) Licensed Operator Requalification Training for EPU (Section 71111.11)

Plant Modifications (all Units) (1 sample)

(1) DCN 69424 Replace Unit 3 Condenser Vacuum Pressure Switches with Pressure

===Transmitters (Section 71111.18)

Post-Maintenance / Post-Modification or Surveillance Tests (Unit 3) (2 samples)===

(1) Surveillance 3-SR-3.1.7.3 Standby Liquid Control System Enriched Sodium Pentaborate

===Solution Concentration, Quantity Calculation, and ATWS Equivalency Calculation following implementation of modified boron enrichment for Extended Power Uprate (Section 71111.19)

(2) 3-SR-3.3.1.1.13 APRM 1-4 calibrations in accordance with DCN 68463 Stage 4 associated with the EPU modification (Section 71111.22)

Regulatory Commitments and Recommended Areas for Inspection (all Units) (1 sample)===

(1) Regulatory Commitments related to EPU (Section 71152)

Identification and Resolution of Problems (Unit 3) (1 sample)

(1) Problem Identification and Resolution related to EPU (Section 71152)

Flow Accelerated Corrosion and Erosion Corrosion Program Reviews (all Units) (2 samples)

(1) Flow Acceleration Corrosion Program (Section 71004)

(2) Erosion Corrosion Program (Section 71004)

71111.19 - Post Maintenance Testing

Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green [H.1] - 71111.19 Systems NCV 05000259, 260, 296/2018001-01 Resources Closed

Introduction:

A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion V, was identified when the licensee failed to perform an adequate post-maintenance test in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the 3C diesel generator output breaker did not ensure that all contacts on replacement stationary switches successfully changed state after installation.

Description:

On February 20, 2018, during the biannual performance of TS SR 3.8.1.9 for the 3C diesel generator, several automatic safety functions did not occur as designed. The 3B RHR, 3B Core Spray, and B1 RHRSW pumps did not automatically start after the 3C diesel generator output breaker closed in to the 3EC 4kV Shutdown Board. The Unit 3 480V Load Shed for Division II also did not occur. The degraded condition was determined to be the result of one pair of contacts on the diesel generator output breakers stationary switch failing to make up when the breaker closed in. Troubleshooting revealed that the stationary switch contact failed to make up because the associated actuating arm on the breaker failed to rotate the stationary switch sufficiently. Although these actuations did not automatically occur, they could have been accomplished manually once recognized by control room operators.

This particular contact was used in a part of the logic circuitry to signify that the diesel generator had successfully tied onto the board and was ready to accept the designed safety loads when there was an accident signal present and normal offsite power to the board was not available. The contact also initiates load shedding of non-essential 480 volt loads to prevent the diesel generator from being overloaded as the safety loads are automatically sequenced on. Additionally, because the 3B Core Spray pump would not have automatically started, the 3D Core Spray pump would also not have automatically started because of the design of the Core Spray initiation logic. The last time that the switch was known to be working correctly was during the last biannual surveillance test in February of 2016. The licensees past operability evaluation concluded that the 3C diesel generator, 3B and 3D Core Spray pumps, 3B RHR pump, B1 RHRSW pump, and Unit 3 480V Division II Load Shed Logic be considered inoperable from February 25, 2016, until February 20, 2018.

From a review of historical maintenance on this breaker, it was identified that the switch was replaced on March 3, 2016, via work order 116872223 as a 24 year preventative maintenance action; however, only a portion of the switchs contacts were tested for continuity during the post-maintenance tests. Inspectors identified that the testing performed did not satisfy the requirements of NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, section 3.2.2.A.5 required that, PMTs for safety-related circuits shall include testing to ensure affected portions of the logic circuitry are tested if they were potentially affected.

Corrective Action(s): The breaker stationary switch was replaced and retested satisfactorily.

Corrective Action Reference(s): CR 1389131

Performance Assessment:

Performance Deficiency: The failure to perform adequate post maintenance testing on the 3C diesel generator output breaker in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing, was a performance deficiency.

Screening: The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency caused the licensee to return a safety-related breaker to service that was later discovered to not be able to perform all of its safety related functions and rendered multiple supported components inoperable.

Significance: Using Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as requiring a detailed risk evaluation because it resulted in an actual loss of function of at least a single train for greater than its TS allowed outage time. An NRC Regional Senior Reactor Analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6 and SPAR Model Version 8.50 for Unit 3. The SRA modeled the condition by assuming the EDG 3C Load Sequencer was failed for one year, which accounted for pump automatic start failures, and that manual start remained available. To account for potential manual start failures, the SRA performed a human reliability analysis using the SPAR-H method and adjusted the model to include a probability of operator failure to recover the sequencer. The dominant sequences (12), which accounted for 90% of the change, involved loss of offsite power with failure of various EDG combinations leading to a station blackout, loss of suppression pool cooling, and failure of low pressure injection. The result was a change in core damage frequency of less than 1E-7/year and was primarily mitigated by operator recovery. Because the change was less than 1E-7/year, no further analysis was needed for external events or large early release, and this finding was determined to be of very low safety significance (Green).

Cross Cutting Aspect: [H.1] - Resources. The apparent cause of the performance deficiency was that leaders did not ensure that plant procedures contained guidance for developing adequate post-maintenance tests for breaker stationary switch replacements.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions Procedures and Drawings, states, in part, that instructions shall include appropriate quantitative and qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, on March 3, 2016, work order 116872223 did not contain post-maintenance test instructions with appropriate acceptance criteria for determining that the breaker stationary switch replacement had been satisfactorily accomplished.

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

71124.01 - Radiological Hazard Assessment and Exposure Controls

Unauthorized Entry into a High Radiation Area (HRA)

Cornerstone Significance Cross-cutting Report Aspect Section Occupational Green [H.8] - 71124.01 Radiation Safety NCV 05000259, 260, 296/2018001-02 Procedure Closed Adherence

Introduction:

A self-revealing, Green, NCV of Technical Specification (TS) 5.7.1, was identified for a worker who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3 A & C Residual Heat Removal (RHR) heat exchanger room using an incorrect Radiation Work Permit (RWP) and without being briefed on the radiological conditions.

Description:

On March 24, 2018, an electrician was assigned the job of installing a jumper on a component in the Unit 3 A & C RHR heat exchanger room. At the time, this area was posted Contaminated Area and High Radiation Area. The electrician logged into RWP 18370011, which did not allow entry into HRAs. The worker also bypassed the Radiation Protection (RP) desk and failed to receive a briefing on radiological conditions in the area.

The worker then dressed in anti-contamination clothing and proceeded past the HRA boundary into the room. He subsequently received a dose rate alarm of 82 mrem/hr, which exceeded the ED alarm setpoint of 60 mrem/hr, and immediately exited the area. A RP technician performed a follow up survey and confirmed the presence of HRA dose rates up to 300 mrem/hr at 30 cm.

Corrective Action(s): The licensee took immediate corrective actions including Radiologically Controlled Area (RCA) access restriction for the individual and initiation of an investigation of the event including surveys of the areas entered.

Corrective Action Reference(s): CR 1390579

Performance Assessment:

Performance Deficiency: The workers entry into a HRA without using an appropriate RWP and without being briefed on radiological conditions in the area, as required by TS 5.7.1, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.

Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter (IMC) 0609 C, Occupational Radiation Safety Significance Determination Process. The finding was not related to As Low As Reasonably Achievable (ALARA)planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green).

Cross-cutting Aspect: This finding involved the cross-cutting aspect of Human Performance, Procedural Adherence, because the event was a direct result of the workers failure to adhere to administrative requirements for HRA access.[H.8]

Enforcement:

Violation: Technical Specification 5.7.1 requires that access to HRAs be controlled by means of an RWP and entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to this, on February 24, 2018, a licensee employee entered a posted high radiation area without proper RWP authorization and without being knowledgeable of the radiological conditions. Upon identification, the licensee immediately implemented RCA access restrictions for the individual and completed follow up surveys of the areas entered.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Implement Controls for Locked High Radiation Area (LHRA) Access Cornerstone Significance Cross-cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV 05000259/260/296/ 2018001-03 Teamwork Closed

Introduction:

A self-revealing, Green, NCV of TS 5.7.2, was identified for the failure to control access to a LHRA. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without RP personnel present. In doing so, the worker accessed an area with dose rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry.

Description:

On February 18, 2018, a carpenter was directed by the RP Drywell Coordinator to install a ladder on the 563 elevation of the Unit 3 drywell near the A blower bank. The inspectors noted the ladder allowed access to an area that had not been surveyed by RP, was not posted or controlled as a LHRA, and no RP technician was present during the installation. While climbing up the ladder to complete a tie off, the carpenter received a dose rate alarm of 458 mrem/hr which exceeded the ED alarm setpoint of 400 mrem/hr. The ED alarm was seen by the remote monitoring station and a roving RP technician was dispatched to respond. The RP technician directed the carpenter to exit the drywell and report to RP.

The technician immediately performed a survey of the area accessible by the ladder and discovered dose rates up to 20 rem/hr on contact and 6 rem/hr at 30cm.

Corrective Action(s): The licensee took immediate corrective actions including posting a LHRA guard until appropriate controls could be implemented.

Corrective Action Reference(s): CR 1388425

Performance Assessment:

Performance Deficiency: The failure to post, lock, and survey the area prior to entry (or be escorted by RP), as required by TS 5.7.2, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.

Significance: The inspectors assessed the significance of the finding using IMC 0609 C, Occupational Radiation Safety Significance Determination Process. The finding was not related to ALARA planning, nor did it involve an overexposure or substantial potential for overexposure (due to the use of remote monitoring), and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green).

Cross-cutting Aspect: This finding involved the cross-cutting aspect of Human Performance, Teamwork, because the event was a direct result of poor coordination between work groups.

[H.4]

Enforcement:

Violation: Technical Specification 5.7.2 requires that HRAs with dose rates > 1 rem/hr at 30 cm, but less than 500 rad/hr at 1 m, be conspicuously posted and provided with a locked or continuously guarded door. TS 5.7.2 also requires that, except for personnel escorted by RP, entry into such areas be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to this, on February 18, 2018, a licensee employee installed a ladder that allowed access to an area with dose rates > 1 rem/hr at 30 cm, but less than 500 rad/hr at 1 m, that was not posted or locked. In addition, the employee entered the area without a RP escort and prior to dose rates being determined.

The licensee took immediate corrective actions including posting a LHRA guard until appropriate controls could be implemented.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

71152 - Problem Identification and Resolution

Inadequate Configuration Control of HPCI Valve Design Issues Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green None

71152 - Annual

Systems NCV 05000296/2018001-04 Follow-up of Closed Selected Issues

Introduction:

A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, was identified when the licensee failed to ensure adequate control of valve design configurations in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control Revision 6. Specifically, the licensee machined a HPCI discharge valve contrary to original design and did not document the change.

Description:

On September 24, 2017, during the performance of the quarterly HPCI pump test an unintentional injection of colder condensate water into the reactor vessel occurred causing reactor power to be at 104% power for about 5 minutes. The injection was caused by a fractured actuator yoke nut that had developed during the June 2017 stroke test of 3-FCV-73-44 leaving the valve partially open. The licensee disassembled and inspected 3-FCV-73-44, and three other valves as a part of their extent of condition review.

During the disassembly of the valves, the licensee identified that the yoke nut flanges on two of the valves were found to be 1 versus that specified in the original vendor drawing which showed the flange was 1.25. The licensees evaluation determined that during past modifications of these valves the yoke nuts were received from the vendor and machined down to 1 without approval or documentation. Licensee extent of condition reviews identified another HPCI valve with an unapproved and undocumented 0.25 spacer below the bottom bearing set. Other deviations identified, were missing ball bearings and additional components in the bearing housing (bearing cage).

Corrective Action(s): As an immediate corrective action the licensee restored each of the valves to their original configurations in accordance with the vendor drawings.

Corrective Action Reference(s): CRs 1341458, 1357076, 1347334, and 1359556

Performance Assessment:

Performance Deficiency: The failure to ensure adequate control of valve design configurations, as required by NPG-SPP-9.3 revision 6, was a performance deficiency.

Specifically, the licensee machined a HPCI discharge valve contrary to original design and did not document the change.

Screening: The performance deficiency was more than minor because it was associated with the design control attribute and affected the associated cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the resulting yoke nut and bearing configuration contributed to the failure of the valve, and prevented the valve from stroking fully closed.

Significance: Using Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a deficiency that affected the design and qualification of safety related, HPCI valves, but operability was maintained.

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

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

NPG-SPP-9.3 establishes a process of administrative controls and regulatory/quality requirements for plant modifications and changes to engineering documents. NPG-SPP-9.3 Rev. 6, Step 3.1.9.A.1 states, in part, that vendor manuals and configuration control design documents affected by the change package have been revised or updated. Contrary to the above, in April 2012, the licensee failed to ensure that vendor manuals and other configuration control design documents affected by the change were revised or updated for 3-FCV-73-44.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

This finding closes URI 05000260, 296/2017008-03 Unresolved Item URI 05000260, 296/2017008-01, Potential

71152 - Annual

(Closed) Inadequate Weak Link Analysis for Unit 2 and Unit Follow-up of 3, HPCI Discharge Valves Selected Issues

Description:

The subject URI was identified to determine if a performance deficiency exists regarding the adequacy of the weak link analysis for the valve and actuator of the HPCI Unit 2 and Unit 3 discharge valves. Inspectors reviewed the various historical weak link analyses for these valves. The original vendor analysis only included the results for the most limiting part in the valve rather than a complete documented analysis for each area analyzed. This minimal level of documentation met the licensees and regulatory standards. As a result, the licensee had no documentation that would cause engineers to believe that the valves yoke nut or yoke nut bearings would exceed their load ratings once the valves actuator thrust was increased in 2012. The valve vendor failed to recognize these loading limitations during their reviews that supported the licensees thrust modification. As a result of this discovery, Crane Nuclear Inc. issued a 10 CFR Part 21 Notification of Defect to the NRC on December 19, 2017.

Corrective Action Reference(s): CR 1344131 Closure Basis: Inspectors concluded that the defects described in the valve vendors notification were not reasonably within the licensees ability to foresee and did not represent a performance deficiency.

Unresolved Item URI 05000296/2017008-02, Potential Inadequate

71152 - Annual

(Closed) Commercial Grade Dedication of Components in Follow-up of Safety Related Valves Selected Issues

Description:

The subject URI was identified to determine if a performance deficiency existed regarding the adequacy of the commercial grade dedication of the valve yoke nut bearings in the HPCI discharge valves on Unit 2 and Unit 3.

Corrective Action Reference(s): CR 1358257 Closure Basis: Since the original thrust bearings were purchased/provided directly from the valve manufacturer, the licensees commercial grade dedication process was not applicable and there was no performance deficiency attributable to the licensee associated with the variation in bearing configuration. The acceptability of the valve manufacturers dedication process for the commercial grade bearings was not within the scope of this inspection.

Replacement bearings were procured after the as-found configurations were discovered to be different than the original design configuration. These replacement bearings were procured as commercial grade items and dedicated by the licensee prior to installation. No findings were identified.

Unresolved Item URI 05000296/2017008-04, Potential Inadequate

71152 - Annual

(Closed) Operator Response to Inadvertent HPCI Injection Follow-up of Selected Issues

Description:

The subject URI was identified to determine if a performance deficiency exists regarding the adequacy of control room operators response to the September 24, 2017, Unit 3 inadvertent HPCI system injection into the reactor vessel.

Prior to the surveillance, reactor power had been reduced to 99.3 percent. The inadvertent injection caused reactor power to exceed the 100 percent licensed thermal power limit (RTP)and initiated an alarm for a reactor feedwater control system input failure. After the alarm, operators noticed that the HPCI check valve 3-73-45 was indicating open despite the upstream discharge valve 3-FCV-73-44 indicating closed. Once the operators diagnosed that HPCI injection was occurring, they initiated a HPCI turbine trip. The HPCI injection lasted approximately five minutes and reactor power stabilized at 104.8 percent. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average RTP was less than 100%.

The inspectors reviewed the licensees performance analysis, Regulatory Information Summary (RIS) 2007-21, Adherence to Licensed Power Limits and IMC 0612, Appendix E, Examples of Minor Issues which discussed this circumstance. The training analysis concluded that the crew did not understand the expected plant response with a HPCI injection and thus were delayed in performing actions specified in AOI-3-1, Loss of Reactor Feedwater. Step 15 directed tripping the HPCI pump. The RIS stated that thermal power may rise slightly due to normal changes in plant parameters and operators are expected to take prompt corrective action to reduce thermal power once it is discovered to be above the licensed limit. Licensees may not intentionally operate or authorize operation above the maximum power level as specified in the license.

IMC0612, Appendix E found this circumstance to be one of minor significance when:

• Operators had performed the prerequisite power reduction and after realizing that thermal power had exceeded RTP, promptly decreased thermal power below the RTP.

• Operators made appropriate and timely adjustments to prevent the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average CTP from exceeding RTP Corrective Action Reference(s): CR 1346991 Closure Basis: The Inspectors concluded that there was no intentional operation above RTP and that the operator response met the guidance in both the RIS 2017-21 and the IMC 0612, Appendix E.

Observation

71152 - Annual Follow-up of Selected Issues

For the implementation of Unit 3 extended power uprate, inspectors assessed the licensees performance regarding problem identification and resolution against selected attributes listed in section 03.06 of Inspection Procedure 71152. Inspectors reviewed condition reports associated with extended power uprate to verify that problems were being promptly identified, evaluated, prioritized and resolved within the licensees corrective action program. Inspectors also reviewed the NRC Safety Evaluation for any regulatory commitments associated with extended power uprate and found that the licensee did not make any regulatory commitments.

Overall, inspectors found no licensee performance weaknesses during this review.

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public disclosure.

• On January 25, 2018, the EC/FAC inspection results were presented to Steve Bono and other members of the licensee staff

• On March 2, 2018, the radiation protection inspection and in-service inspection results were presented to Mr. D. L. Hughes and other members of the licensee staff.

• On April 20, 2018, the quarterly resident inspector inspection results were presented to Mr.

Werner Paulhardt and other members of the licensee staff.

DOCUMENTS REVIEWED

IP 71111.04

Procedures

3-OI-74, Residual Heat Removal System, Revision 125

0-OI-72, Auxiliary Decay Heat Removal System, Revision 60

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

Drawings

3-47E811-1, Flow Diagram Residual Heat Removal System, Revision 72

0-47E873-1, Flow Diagram Aux Heat Removal System, Revision 8

0-15E740-1, Single-Line Diagram ADHR Service Entrance and MCC, Revision 13

Other Documents

CR 1388305

0-BFN-VM-5003, Installation, Operation and Maintenance Instructions and Engineering

Document Package

FSAR Chapter 8.4, Normal Auxiliary Power System

DCN 71673, Implementation of U3 Emergency High Pressure Make-Up Pump System

IP 71111.05

Procedures

Browns Ferry Fire Protection Report-VOLUME 2, Fire Protection Report Volume 2, Revision 58

Other Documents

MDN0009992012000100, Browns Ferry Nuclear Power Plant, Units 1, 2, and 3, Fire Risk

Evaluations, Revision 6

EDQ099920110010, NFPA 805 - Nuclear Safety Capability Analysis, Revision 33

IP 71111.06

Drawings

2-47W2392-642, Fire Protection - 10CFR50 Appendix R Penetration Seal Tabular Drawings

E

L. 621.25, Revision 2

0-47W510-1, Mechanical Roof Drains, Revision 1

0-47W510-2, Mechanical Roof Drains, Revision 4

Other Documents

BFN-57250, BFN-0-PMP-040-0031, Visual Inspection of Listed Handholes and Sumps Per

95003 Commitment, Revision 6

WO 118861289

CR 1375311

CR 1375316

NDN-000-999-2007-0031, Internal Flooding BFN Probabilistic Risk Assessment, Revision 0

DED-TM-PF2, Concluding Report of the Effects of Postulated Pipe Failure Outside of

Containment for the Browns Ferry Nuclear Plant Unit s 2 and 3, dated March 1, 1974

IP 71111.08

Procedures

N-UT-64, Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds, Revision

0016

N-UT-78, PDI Generic Procedure for the Manual Ultrasonic Examination of Reactor Pressure

Vessel Welds, PDI-UT-6, Revision 9

N-UT-90, Generic Procedure for the Ultrasonic Detection and Sizing of Reactor Pressure Vessel

Nozzle to Shell Welds and Nozzle Inner Radius, Revision 003

N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Revision

0047

PDI-UT-2, PDI Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds,

Revision H, October 4, 2017

PDI-UT-6, PDI Generic Procedure for the Ultrasonic Examination of Reactor Pressure Vessel

Welds, Revision I, August 1, 2017 PDI-UT-11, Generic Procedure for the Ultrasonic

Examination of Reactor Pressure Vessel Nozzle-to-Shell Welds and the Nozzle Inner Corner

Radius, Revision D 08-01-2017, Revision D, August 1, 2017

Drawings

2-47W2392-6

Other Documents

CDQ0-003-2006-0002, BFN RPV Feedwater Nozzles Fatigue and Fracture Mechanics

Evaluation, Revision 14

CR 1135166, ISI Examination Drawings

CR 1143845, CR to Track Accept-As-Is for Indication on Top of Unit 3 Vessel Head

CR 1145011, FME Voluntary Stop Work for 3A Feed Water Heater Welding

CR 1145022, FME Procedure Not Followed by Contractors

CR 1145738, Incorrect Detail Weld Procedure Specification

CR 1146291, Documentation Errors on Weld Data Sheets

CR 1146995, Tack Welds Made without Sufficient Purging

CR 1146888, Potential Rework Event

CR 1147745, Discrepancies and Errors on Weld Data Sheet

CR 1147756, A D&Z Mods Welder Contaminated in RCA Clean Area

CR 1148490, U3R17 Jet Pump Wedge Wear and Set Screw Gaps / Indications

CR 1150215, Welding Being Performed without a Fire Watch

CR 1150705, NOI U3R17-007: Moisture Seal Barrier (MSB) Loss of Adhesion.

CR 1166944, Core Shroud Off-Axis Cracking Interim Inspection & Flaw Evaluation Guidance

CR 1184618, Through-Wall Penetration in Safety-Related Heat Exchanger Shell

CR 1187114, Part 21 - Inadequate Vendor Documentation of Far Vision Acuity Certifications

CR 1210910, Potential Code Class-2 Piping Leak on RBCW Piping @ 1-DRV-70-507

Connection Elbow

CR 1221309, Two Welding Machines Left On and Unattended

CR 1223258, Invertec V350 Pro Welder Left On and Unattended

CR 1227532, Scheduled Containment ISI Examination Not Performed

CR 1229969, Leakage Coming from 1-CKV-73-45

CR 1244822, Welding Sparks Escaped Containment Tent on RFF

CR 1250683, Request for Review of BWRVIP Position Regarding Aging Management of

Orificed Fuel Support Castings

CR 1284288, Re-Welding Stainless Steel Multiple Times Presents Various Issues

CR 1324316, (CRP-ENG-FSA-17-004) ISI Program Deficiencies

CR 1326645, BWRVIP Skip Outage Project Initiation

CR 1333664, BFN Leak Source Evaluation

Browns Ferry Nuclear Standard ISI Plan (Baseline) Standard Code ASME Section XI, 2007 Ed /

2008 Add Category Scheduling Compliance

Calibration Block WB-084 As-Built Verification Documentation

Certification for Magnaflux Ultragel II, Batch Number 16H031

Certificate of Compliance for Miniature Angle Beam Block, Serial Number 789631

Certificate of Compliance for Miniature Angle Beam Block, Serial Number 791719

Certificate of Conformity I07120001 for Visual Illumination Cards

Certified Material Test Reports for weld rods used for WOs 117544712 and 117656145

CRP-ENG-FSA-17-004, Focused Self-Assessment Report, Inservice Inspection at Browns

Ferry, Approved September 14, 2017

Detail Welding Procedure Specification (DWPS) GT88-O-1-N, Manual Gas Tungsten Arc

Welding, Revision 5

Drawing 3-47B400-99, Mechanical, Main Steam System Pipe Support, Revision 000

Drawing BF-18, Calibration Blocks As-Builts BF-18, Material: A-533, Revision 01

IVVI Examination Checklist Browns Ferry Unit 3 R18 Spring 2018 (BF3R18) Outage

Krautkramer Transducer Certification for Number 01FH9V

Krautkramer Transducer Certification for Number 16B003AA

Krautkramer Transducer Certification for Number 16B003AC

Krautkramer Transducer Certification for Number 16B003AG

Letter to TVA from NRC, dated March 14, 2017, Subject: Browns Ferry Nuclear Plant, Units 2

and 3 - Request for ASME Code,Section XI, Alternatives 2-ISl-30 and 3-ISl-27 for the Periods

of Extended Operation Regarding Reactor Pressure Vessel Circumferential Shell Weld

Examinations

Owners Activity Report for BFN, Unit 3, Cycle 17 Operation, dated 6/21/16

NDE Personnel Qualifications for

J. Hoover,

M. Kleinjan,

D. Maclean, D. Sawatzky

Report of Calibration for Krautkramer Ultrasonic Flaw Detector, Serial Number 0100H4

Report of Calibration for Krautkramer Ultrasonic Flaw Detector, Serial Number 0132M6

Report of Calibration for TEGAM Digital Thermometer, Serial Number T-257196

Report of Calibration for Keithley Digital Thermometer, Serial Number T-12463

UT Examination Report R-049, Pipe to Elbow Weld, Component ID: DSRHR-3-04

UT Examination Report R-085, Nozzle to Shell Weld, Component ID: N3D-NV

VT Examination Report R-033, Pipe Support, Component ID: 3-478400-099

VT Examination Report R-069, Variable Spring Can, Component ID: RHR-3-H-1

Welder Qualification Records for

C. Brock,

K. Davenport,

J. Gautney,

C. Hill,

C. Lindsey,

J. Parker,

S. Laird, and E. Woods

Weld Map and Data Sheets for WOs 117544712 and 117656145

Welding Procedure Qualification Record GTA 88-0-1, Gas Tungsten Arc Welding, dated

December 29, 1978

Welding Procedure Qualification Record GTA 88-0-5, Gas Tungsten Arc Welding, dated

April 15, 2004

WO 117544712, HPCI Mod per DCN 71865, Valve 73-23 and 73-603 to be Relocated

WO 117656145, Replace Valve BFN-3-TV-069-0583

IP 71111.11

Procedures

3-AOI-3-1, Loss of Reactor Feedwater or Reactor Water Level High/Low, Revision 12

3-AOI-1-1, Relief Valve Stuck Open, Revision 14

NPG-SPP-17.8.4, Conduct of Simulator Operations, Revision 4

BFN-ODM-4.20, Strategies for Successful Transient Mitigation, Revision 4

3-GOI-100-1A, Unit Startup, Revision 116

0-TI-248, Station Reactor Engineer, Revision 113

NPG-SPP-10.4, Reactivity Management Program, Revision 6

3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in

Power During Power Operations, Revision 61

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

3-OI-47, Turbine-Generator System, Revision 11

1-GOI-100-1A, Unit Startup, Revision 48

Other Documents

OPL175S055, SRV Fails Open, HPCI inadvertent actuation, 3B 4kV Unit Board Trip, ATWS with

MSIVs Open, Revision 0

IP 71111.12

Procedures

0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -

10CFR50.65, Revision 50

Other Documents

System Health Report for System 575 4kV AC Power Distribution,

U1/2&3 Function 575-B, C & E 4kV Power Supply Busses Sys (a)(1) Plan, Revision 11,

Effective October 27, 2017

Functional failure and Unavailability data for System 575 through February 2018

IP 71111.13

Procedures:

BFN-ODM-4.18 Protected Equipment, Revision 17

NPG-SPP-09.11.1 Equipment Out of Service Management, Revision 12

0-TI-248, Reactor Engineer, Revision 113

3-OI-47, Turbine-Generator System, Revision 111

MSI-0-000-LFT001, Lifting instructions for the control of heavy loads, Revision 0074

FSAR Appendix C, Structural Qualifications of Subsystems and Components, C.8, Control of

Heavy Loads

1-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Revision 8

3-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Revision 0017

NPG-SPP-10.6, Infrequently Performed Test or Evolutions, Revision 1

MCI-0-085-CRD001, Control Rod Drive Removal and Installation, Revision 0061

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

Drawings:

3-47E610-47-1, Mechanical Control Diagram Turbine, Revision 6

3-47E610-47-4, Mechanical Control Diagram Turbine, Revision 21

3-9952-582, Diagram Main Turbine Control Wiring, Revision A

Other Documents:

CR 1292238

Operator logs from May 4, 2017 through May 5, 2017

Protected equipment list May 05, 2017

Equipment Apparent Cause Evaluation for PER 959856

CR 1379519

Clearance 3-TO-2018-0001 Section 3-001-0004

OPL171.228, Electro-Hydraulic Control Logic, Revision 6

OPL171.230, Electro-Hydraulic Control (EHC) System: Hydraulics, Revision 4

OPL171.010, Main Turbine, Revision 13

ESG116.001C, Electro-Hydraulic Control (EHC) System, Revision 0

FSAR Chapter 7.11, Pressure Regulator and Turbine-Generator Control

FSAR Chapter 11.2, Turbine Generator

50.59 package for CR 1379519

ODMI for CR 1379519

Unit 3 Cycle R18 Outage Safety Plan, Revision 0

IP 71111.15

Procedures:

OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking,

Revision 24

3-SR-3.8.1.9(3C), Diesel Generator 3C Emergency Load Acceptance Test, Revision 23

0-AOI-57-1A, Loss of Offsite Power (161 and 500KV)/Station Blackout, Revision 107

Drawings:

3-47E811-1, Flow Diagram Residual Heat Removal System, Revision 72

Other Documents:

CR 1344119

PDO for CR 1344119

PDO for CR 1349343

CR 1341458 Level 1 Evaluation (RCA) Report

CR 13799519

FSAR Chapter 14.10.1, Events Resulting in a Nuclear System Pressure Increase

TS Bases 3.3.1.1

ODMI for CR 13799519

OPL171.228, Electro-Hydraulic Control Logic, Revision 6

OPL171.230, Electro-Hydraulic Control (EHC) System: Hydraulics, Revision 4

NDQ0074880118, Evaluation of LPCI Flow to Reactor Pressure Vessel (RPV) with Failed Open

Min-Flow Bypass Valve, Revision 6

MDQ0074920028, System Requirements Calculation for Residual Heat Removal (RHR)

System, Revision 6

FSAR Table 6.5-2, ECCS Equipment Capacity Assumed in LOCA Analysis

NDQ099920100006, Diesel Frequency Variation Evaluation, Revision 0

MDQ0074920113, Documentation of RHR Pump Discharge Test Flow Rates and System Test

Pressure, Revision 0

GE Letter BF 3-7413, Long Term Containment Cooling Requirements - Browns Ferry Unit 3,

dated February 27, 1976

TVA Letter, Additional Information Requested by NRC Concerning RHR Pump Protection

Against Operation in Excess of Design Runout, dated July 21, 2016

CR 1382124

BFN-VTD-G080-0771, Operation and Maintenance Instructions for General Electric NUMAC

Two-Out-of-Four Logic Module, Revision 1

CR 1382150

Browns Ferry Unit 3 Cycle 18 Core Operating Limits Report, Revision 2

ANP-3413P, Browns Ferry Unit 3 Cycle 18 Plant Parameters Document, Revision 0

Past Operability Evaluation for CR 1389131

IP 71111.18

Procedures

3-SIMI-47B, Electro-Hydraulic Control System Scaling and Setpoint Documents, Revision 41

SII-3-XX-47-204.3, Electro-Hydraulic Control System Condenser Vacuum (Turbine Exhaust)

Transmitter Calibration and Functional Test, Revision 0

Other Documents

DCN 69424, Condenser Vacuum Pressure Switches, Revision A

DCN 72342, Modify EHC Software, Revision A

PMTI-72342-03, Install Condenser Vacuum Transmitters and Provide Power Dependent Trip

Signals, Revision 2

BFN-VTD-G080-3095, General Electric Instructions - Allowable Exhaust Pressure Operation,

Revision 2

Turbine Backpressure Evaluation dated May 2016

CR 1399171

IP 71111.19

Procedures

3-SR-3.3.8.1.3 (3EB), Unit 3 4kV Shutdown Board 3EB Loss of Power Logic System Functional

Test. Revision 0008

ECI-0-000-MOV009, Testing of Motor Operated Valves, Revision 46

CCI-0-XI-00-019, Electrical Indicators, Revision 13

BFN-3-INVT-256-0001, Replace ECCS Inverter, PM Job Plan 500126964, Revs. 1, 2

ECI-0-000-BKR008, Testing and Troubleshooting of Molded case Circuit Breakers and Motor

Starter Overload Relays, Revision 0107

3-SI-4.7.A.2.g-3/3a, Primary Containment Local Leak Rate Test Reactor Feedwater Line A:

Penetration X-9A

PMTI-71673-001, Emergency High Pressure Make-up Pump Testing

WO Instructions BFN-3-BKR-211-03EC/012 Test MJ (52 Aux Switch) Switch Normal Feeder

Breaker 1338

3-SR-3.6.1.3.5(RHR II), RHR System MOV Operability Loop II, Revision 32

ECI-0-000-MOV009, Testing of Motor Operated Valves Using Viper 20, Revision 44, 47

0-TI-579(MOV), Motor Operated Valve Data Evaluations, Revision 0

3-SR-3.1.7.3, Standby Liquid Control System Enriched Sodium Pentaborate Solution

Concentration, Quantity Calculation, and ATWS Equivalency Calculation, Revision 50

3-SI-3.1.7.6, Standby Liquid Control System ATWS Equivalency Calculation for Newly

Established Pump Flow Rate, Revision 1

Drawings

3-45E766-18, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E768-3, Emergency Equipment Schematic Diagram Diesel Generator 3B

3-45E766-21, Wiring Diagram 4160V Shutdown Auxiliary Power Schematic Diagram

3-45E766-24, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E766-3, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E724-7, Wiring Diagram 4160V Shutdown BD 3EB Single Line

3-45E68-4, Emergency Equipment Schematic Diagrams

10-11-748, Outline 5KVA inverter 250 VDC 120VAC 10 60HZ

20-113501, Schematic 5KVA Inverter 250VDC 120VAC 10 60HZ

Other Documents

WO 118373618, 118850673, 119428686, 119455067, 118369003, 119460702

IST Evaluation 18-3-IST-074-672 dated March 23, 2018

MDQ3074920442, MOV 3-FCV-74-73, Operator Requirements and Capabilities, Revision 7

IST Evaluation 18-3-IST-073-669 dated March 11, 2018 MDQ3073920407, MOV 3-FCV-73-2,

Operator Requirements and Capabilities, Revision 8

CR 1307478, 1257769, 9563525

Apparent Cause Evaluation Report for PER 956352, Revision 1

MDQ0063900083, Standby Liquid Control System Flow Analysis for ATWS Requirements,

Revision 8

MDQ0063920470, Standby Liquid Control System - Boron 10 Requirements, Revision 5

IP 71111.20

Procedures

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

NPG-SPP-03.21, Fatigue Rule and Work Hour Limits, Revision 20

NPG-SPP-14.1, Fitness-For-Duty and Fatigue Management, Revision 16

3-GOI-200-2A Primary Containment Entry, Revision 3

3-OI-74, RHR System Checklists for Heatup, Initiation of and Loss of Shutdown Cooling

3-GOI-100-3A, Refueling Operations (RX Vessel Disassembly and Floodup), Rev 68

Other Documents

Unit 3, Cycle R18, Outage Safety Plan

IP 71111.22

Procedures

3-SR-3.8.1.9, (3B OL) Unit 3 EDG load acceptance test,

3-SR-3.8.1.9(3B OL) - DG 3B Emergency Load Acceptance Test with Unit 3 Operating,

1-SR-3.5.1.6 (RHR II) Quarterly RHR System Rated Flow Test Loop II

3-SR-3.1.7.7 SLC System Functional Test-Pump

MCI-0-063-VLV001, Maintenance of Fired and Non-Fired SLC Explosive-Actuated Valve Units,

Revision 0025

3-SIMI-92B, Neutron Monitoring System Scaling and Setpoint Documents, Revision 0016

NESSD 3A-092-0001-00-06, Site Engineering Setpoint and Scaling Document Cover Sheet

3-SR-3.3.8.1.3 (3EB) 4KV Shutdown Board 3EB Loss of Power (LOP) Logic System Functional

Test, Revision 8

Other Documents

CR 1372616

WO 118491281

WO 118099436

WO 118491353

IP 71124.01

Procedures

NISP-RP-2, Radiation and Contamination Surveys, Revision 0

NISP-RP-4, Radiological Posting and Labeling, Revision 0

NISP-RP-5, Access Controls for High Radiation Areas, Revision 0

NPG-SPP-05.1, Radiological Controls, Revision 9

NPG-SPP-22.300, Corrective Action Program, Revision 10

RCDP-17, Radiological Postings, Revision 1

RCI-1.2, Radiation, Contamination, and Airborne Surveys, Revision 37

RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Revision 89

RCI-33, Diving Operations on the Refuel Floor, Revision 15

Data

RWP 17220021, U2 FIN Maintenance Activities

RWP 18390039, U3R18 Refuel Floor Dryer Replacement Activities

RWP 18380142, U3R18 Drywell Undervessel Maintenance Activities

RWP 18380032, U3R18 Drywell Carpenter Support Activities, Revision 1

RWP 18370011, U3R18 Reactor Building General Maintenance, Revision 0

Radiological Survey M-20180218-31, U3 Drywell 563

Radiological Survey M-20180219-18, U3 Drywell 563 Posted Ladder Access to LHRA on Top

Radiological Survey M-20180225-7, U3 RXB 565 A & C RHR Hx Update Survey

Radiological Survey M-20180220-22, U3 RXB 519 Under Torus Travel Path Survey

Radiological Survey M-20180224-22, U3 RXB 565 General Area

Radiological Survey M-20180227-56, U3 RXB 565 General Area

Radiological Survey M-20180228-30, U3 DW 550 Sub-pile Room

Radiological Survey M-20180228-15, U3 DW 550 Sub-pile Room

Radiological Survey M-20170919-20, U2 RXB 593 RWCU Pump Room

Radiological Survey M-20170922-2, U2 RXB 593 RWCU Pump Room

Radiological Survey M-20180225-14, U3 RXB 664 Refuel Floor Steam Dryer Diving Activities

Air Sample Record 18-20166-5, U3 Drywell Under Vessel GA

Air Sample Record 18-20183, U3 Drywell 550 Subpile GA

Air Sample Record M-20180301-39, 550 Subpile Room

Air Sample Record 18-20171-6, U3 Refuel Floor Divers Station

Air Sample Record 18-20107, U1 RXB 593 1A RWCU Pump Room

Six Month Inventory and/or Leak Test, August 3, 2017

U3 non-fuel inventory of Spent fuel Pool, July 14, 2017

ALARA Plan 18-0030, U3R18 Outage Steam Dryer Replacement

Tritium Activity Worksheet, U3 RCS, February 20, 2018

Alpha Level 2 and 3 Data Spreadsheet

CAP Documents

Self-Assessment BFN-RP-SSA-18-001

CR 1329077

CR 1326911

CR 1388425

CR 1390579

CR 1388749

CR 1283906

CR 1296485

CR 1291073

CR 1287739

CR 1295731

CR 1329533

IP 71124.08

Procedures, Guidance Documents, and Manuals

NPG-SPP-05.9.1, Radioactive Material/Waste Shipments, Revision 4

NPG-SPP-22.000, Performance Improvement Program, Revision6

NPG-SPP-22.300, Corrective Action Program, Revision 10

RCI-43, Radioactive Material Control, Revision 10

RWI-001, Administration of the Radioactive Material and Radwaste Packaging and

Transportation Program, Revision 12

RWI-005, Radwaste Routines, Revision 13

RWI-111, Storage of Radioactive Waste and Materials, Revision 24

RWI-156, Packaging Radioactive Material and Radioactive Waste, Revision 2

RWTP-101, 10 CFR 61 Waste Characterization, Revision 2

RWTP-102, Use of Casks, Revision 2

0-PCP-001, Process Control Program Manual (PCP), Revision 4

Records and Data

Certificate of Completion, Energy Solutions DOT/NRC Radioactive Waste Packaging,

Transportation and Disposal Training [for three qualified shippers], Various Dates

Design Change, DCN 72581, Activate LLRW Trash Module 1 to Allow Storage of Old Steam

Dryers from EPU Modifications, September 28, 2017

List of Design Changes and Temporary Modifications/ Alterations of Radwaste System since

March 1, 2016, Undated

List of Abandoned Rad Waste Processing Equipment, Undated

Radiological Survey # M-20180216-14, 10CFR37 LLRW Yard Rad Material Storage Area

Update, February 16, 2018

RCI-43, Att. 4 List of Radioactive Material Storage Areas Located Outside the Main RCA,

February 23, 2018

Radioactive Shipment Logs, Browns Ferry Nuclear Plant Radioactive Material Shipment Logs

for Calendar Years 2016, 2017, and 2018 thru January, Various

CFR Part 61 Data Waste Stream Characterization - Scaling Factors for CWPS 2016,

April 17, 2018

CFR Part 61 Data Scaling Factor Analysis-DAW 2016, April 17, 2016

CFR Part 61 Data Scaling Factor Analysis-RWCU 2016, April 17, 2016 10 CFR Part 61 Data

Scaling Factor Analysis-Thermex 2016, April 17, 2016

CFR Part 61 Data Waste Stream Characterization - Scaling Factors for CWPS 2017,

June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-DAW 2017, June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-RWCU 2017, June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-Thermex 2017, June 8, 2017

Shipping Records

Shipment ID # 160107, NCMD Samples (Type A), January 3, 2016

Shipment ID # 160814, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

August 25, 2016

Shipment ID # 170201, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

February 3, 2017

Shipment ID # 171109, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

November 16, 2017

Shipment ID # 171213, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

December 15, 2017

Corrective Action Program (CAP) Documents

CR 1147306, 1170976, 1184022, 1219274, 1248641, 1249712, 1255883, 1307846

Self-Assessment, BFN-RP-SSA-18-001, Radiation Hazards and Transportation

December 12, 2017

IP 71151

Procedures

NPG-SPP-02.2, Performance Indicator Program, Revision 10

Desktop Guide for Identification and Reporting of NEI 99-02 Performance Indicators for

Occupational Exposure Control Effectiveness

Other Documents

Electronic Dosimeter Alarm Report, April 1, 2017 - February 12, 2018

Reactor Coolant System Leakage logs from January 1, 2017 to December 31, 2017

CR 1344430, 1330667, 1356696

IP 71152

Procedures

NEDP-8.2, Technical Evaluation for Procurement of Safety Related and Quality Related

Materials, Items, and Services, Revision 2

Other Documents

CR 1387156, 1384874, 1381298, 1375811, 1370091, 1287517, 1273615, 1262776, 1209499,

1171982, 1139533, 1133786, 1130256, 1088344, 1049856, 1038699, 1007813, 993114,

988512, 985013, 984499

Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Extended Power

Uprate Related to Amendment Nos. 299, 323, and 283 to Renewed Facility Operating License

Nos. DPR-33, DPR-52, and DPR-68 Tennessee Valley Authority Browns Ferry Nuclear Plant,

Units 1, 2, and 3 Docket Nos. 50-256, 50-260, and 50-296

WO 111044065

CR 1341458, 1357076, 1347334, 1359556

Level 1 Evaluation (RCA) Report for CR 1341458

PEG Package CYJ557C-UPGR

TVA Central Lab Services Technical Report AU27033

PEG Package 1781269-BFNX0

CNI Corrective Action Report 17-33

IP 71153

Procedures

3-AOI-100-1, Reactor Scram, Revision 65

3-AOI-100-1, Attachment 1, Scram Report dated January 10, 2018

Drawings

3-47E610-47-1, Mechanical Control Diagram Turbine, Revision 6

3-47E610-47-4, Mechanical Control Diagram Turbine, Revision 21

Other Documents

NRC Inspection Report 2017-001, Section 1R15

CR 1265552

IP 71004

Procedures

0-TI-140, Monitoring Program for Flow Accelerated Corrosion, Revision 7

IEP-200, Qualification and Certification Requirements For TVA Inspection Services

Organization (ISO) Nondestructive Examination (NDE) Personnel, Revision 16

N-UT-26, Ultrasonic Examination For Wall Thinning Conditions, Revision 30

NPG-SPP-09.7.2, Flow Accelerated Corrosion Control Program, Revision 3

Drawings

1-47E801-2, Flow Diagram Main Steam, Revision 5

1-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 19

1-47E804-1, Flow Diagram Condensate, Revision 26

1-47E805-1-1, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 5

1-47E805-1-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 2

1-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 5

2-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 0

2-47E804-1, Flow Diagram Condensate, Revision C

2-47E805-1, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

2-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

3-47E802-1, Flow Diagram Extraction Steam, Revision 0

3-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 0

3-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

Other Documents

BFN-ENG-SSA-18-001, BFN Flow Accelerated Corrosion (FAC) NRC Inspection Readiness

Self-Assessment, dated October 10, 2017

DS-M4.2.1, Flow Accelerated Corrosion Program Methods, Revision 10

EP-2015-0033-01-TR, Browns Ferry Nuclear Plant Unit 1 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-02-TR, Browns Ferry Nuclear Plant Unit 1 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 1

EP-2015-0033-03-TR, Browns Ferry Nuclear Plant Unit 2 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-04-TR, Browns Ferry Nuclear Plant Unit 2 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 0

EP-2015-0033-05-TR, Browns Ferry Nuclear Plant Unit 3 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-06-TR, Browns Ferry Nuclear Plant Unit 3 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 0

Letter 15-0195-LR-001, Letter Report - Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Summary Tables for the Effect of Extended Power Uprate on Flow Accelerated Corrosion,

Revision 0

NCO 040006083, Commitment Completion Form dated October 1, 2012

EPRI State of the Fleet Assessment Tennessee Valley Authority - Browns Ferry Nuclear Plant

dated November 5, 2015

TVA Flow Accelerated Corrosion (FAC) Fleet SelfAssessment dated 6/2017

CR 974071, This SR is to document gaps found in the attached U1 FAC CHECWORKS review

CR 1081644, Evaluate Flow Accelerated Corrosion (FAC) OE from Davis-Besse (INPO Event

Report Level 3)

CR 1115734, A leak was found on the 2C3 FW Heater immediately following the December

2015 outage

CR 1219434, Through wall leak found on elbow downstream of 1-FCV-5-71 valve

CR 1380327, NRC Identified: Evaluate FAC Program SNM Technical Reports for applicability of

S3 susceptibility

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303-1257

May 8, 2018

Mr. J. W. Shea

Vice President, Nuclear Regulatory

Affairs and Support Services

Tennessee Valley Authority

1101 Market Street, LP 4A

Chattanooga, TN 37402-2801

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION

REPORT 05000259/2018001, 05000260/2018001, AND 05000296/2018001

Dear Mr. Shea:

On March 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at your Browns Ferry Nuclear Plant, Units 1, 2, and 3. On April 20, 2018, the NRC inspectors

discussed the results of this inspection with Mr. W. Paulhardt and other members of your staff.

The results of this inspection are documented in the enclosed report.

NRC inspectors documented four findings which were determined to be of very low safety

significance (Green) in this report. All of these findings involved violations of NRC

requirements. Because of their very low safety significance, the NRC is treating these violations

as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest any of the violations or significance of these NCVs, 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,

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; 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.

J. Shea 2

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 10 CFR 2.390, Public Inspections, Exemptions, Requests for

Withholding.

Sincerely,

/RA/

Anthony

D. Masters, Chief

Reactor Projects Branch 5

Division of Reactor Projects

Docket Nos.: 50-259, 50-260, 50-296

License Nos.: DPR-33, DPR-52, DPR-68

Enclosure:

NRC IIR 05000259/2018001,

05000260/2018001 and 05000296/2018001

cc w/encl. Distribution via ListServ

ML18128A153

OFFICE RII/DRP RII/DRP RII/DRP RII/DRP RII/DRS RII/DRS

NAME DDumbacher MKirk ARuh JSeat ANielsen RKellner

DATE 4/27/2018 4/26/2018 4/30/2018 4/27/2018 4/26/2018 4/26/2018

OFFICE RII/DRS RII/DRS RII/DRS RII/DRP RII/DRP RII/DRP

NAME RCarrion RWilliams CCrespo SMonarque PHeher SNinh

DATE 4/26/2018 4/26/2018 4/26/2018 4/27/2018 4/26/2018 5/1/2018

OFFICE RII/DRP

NAME AMasters

DATE 5/8/2018

U.S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-259, 50-260, and 50-296

License Nos.: DPR-33, DPR-52, and DPR-68

Report No.: 05000259/2018001, 05000260/2018001, and 05000296/2018001

Enterprise Identifier: I-2018-001-0052

Licensee: Tennessee Valley Authority (TVA)

Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3

Location: Corner of Shaw and Nuclear Plant Road

Athens, AL 35611

Dates: January 1, 2018 through March 31, 2018

Inspectors:

D. Dumbacher, Senior Resident Inspector

M. Kirk, Resident Inspector

A. Ruh, Resident Inspector

A. Nielsen, Senior Health Physicist

R. Kellner, Senior Health Physicist

R. Carrion, Senior Reactor Inspector

S. Monarque, Project Engineer

J. Seat, Project Engineer

P. Heher, Project Engineer

R. Williams, Senior Reactor Inspector

G. Crespo, Senior Construction Inspector

Approved by:

A. Masters, Chief

Reactor Projects Branch 5

Division of Reactor Projects

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance

by conducting quarterly integrated baseline inspections at Browns Ferry Nuclear Plant, Units 1,

2, and 3 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. NRC and

self-revealed findings, violations, and additional items are summarized in the table below.

List of Findings and Violations

Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches

Cornerstone Significance Cross-cutting Report

Aspect Section

Mitigating Green [H.1] - 71111.19

Systems Non-cited Violation (NCV) 05000259, Resources

260, 296/2018001-01

Closed

A self-revealing, Green, NCV of 10 CFR Part 50 Appendix B, Criterion V, was identified when

the licensee failed to perform an adequate post-maintenance test in accordance with NPG-

SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the

3C diesel generator output breaker did not ensure that all contacts on replacement stationary

switches successfully changed state after installation.

Unauthorized Entry into a High Radiation Area (HRA)

Cornerstone Significance Cross-cutting Report

Aspect Section

Occupational Green [H.8] - 71124.01

Radiation Safety NCV 05000259, 260, 296/2018001-02 Procedure

Closed Adherence

A self-revealing, Green, NCV of Technical Specification (TS) 5.7.1, was identified for a worker

who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3

A & C Residual Heat Removal Heat Exchanger Room using an incorrect Radiation Work

Permit and without being briefed on the radiological conditions.

Failure to Implement Controls for Locked High Radiation Area (LHRA) Access

Cornerstone Significance Cross-cutting Report

Aspect Section

Occupational Green [H.4] - 71124.01

Radiation Safety NCV 05000259/260/296/ 2018001-03 Teamwork

Opened/Closed

A self-revealing, Green, NCV of TS 5.7.2, was identified for the failure to control access to a

LHR

A. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without

Radiological Personnel (RP) present. In doing so, the worker accessed an area with dose

rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry.

Inadequate Configuration Control of High Pressure Coolant Injection (HPCI) Valve Design

Issues

Cornerstone Significance Cross-cutting Report Section

Aspect

Mitigating Green None 71152 - Annual

Systems NCV 05000296/2018001-04 Follow-up of

Closed Selected Issues

A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, was identified

when the licensee failed to ensure adequate control of valve design configurations in

accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control

Revision 6. Specifically, the licensee changed, over time, HPCI discharge valve yoke nut and

bearing components contrary to original design without documenting or evaluating the

changes

Additional Tracking Items

Type Issue Number Title Report Status

Section

URI 05000260,296/ Potential Inadequate Weak Link Analysis 71152 Closed

2017008-01 for Unit 2 and Unit 3, HPCI Discharge

Valves

URI 05000260,296/ Potential Inadequate Commercial Grade 71152 Closed

2017008-02 Dedication of Components in Safety

Related Valves

URI 05000260,296/ Potential Inadequate Configuration Control 71152 Closed

2017008-03 of the Unit 2 and Unit 3 HPCI Discharge

Valves

URI 05000296/ Potential Inadequate Operator Response 71152 Closed

2017008-04 to Inadvertent HPCI Injection

LER 05000260/2017- Inoperable Primary Containment Isolation 71153 Closed

2-00 Valve Resulting in Condition Prohibited by

Technical Specifications

LER 05000259/2016- Incorrect Tap Settings for 480 Volt 71153 Closed

004-01 Shutdown Transformer Results in

Inoperability of Associated 480V Shutdown

Boards

PLANT STATUS

Unit 1 operated at 100% rated thermal power (RTP) except for a reactor scram related to a

turbine control valve partial closure transient on March 18, 2018. The unit returned to 100%

RTP on March 24, 2018, and operated at that level for the remainder of the inspection period.

Unit 2 operated at 100% RTP for the duration of the inspection period.

Unit 3 operated at 100% RTP until a reactor scram occurred on January 10, 2018, related to

vibration-induced failure of hydraulic piping for the #2 turbine control valve. The unit returned to

100% power on January 15, 2018, and operated at that level until February 17, 2018. There

were two unplanned downpowers during the inspection period due to #3 turbine control valve

oscillations, one planned downpower for 3C reactor feed pump maintenance. From

February 17, 2018, through March 31, 2018, Unit 3 was shutdown for a planned refueling

outage U3R18.

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 reviewed selected procedures and records,

observed activities, and interviewed personnel to assess licensee performance and compliance

with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

71111.01 - Adverse Weather Protection

Seasonal Extreme Weather (1 Sample)

The inspectors evaluated readiness for seasonal extreme weather conditions prior to the

onset of seasonal cold temperatures on January 2, 2018.

71111.04 - Equipment Alignment

Partial Walkdown (4 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following

systems/trains:

(1) Unit 3 Residual Heat Removal (RHR) Loop II Shutdown Cooling alignment on February

18, 2018

(2) Unit 3 Alternate Decay Heat Removal (ADHR) on February 23, 2018

(3) 4160V AC Electrical System on March 10, 2018

(4) Unit 3 Main Steam System on March 15 and 19, 2018

Complete Walkdown (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit

Emergency High Pressure Makeup (EHPM) system on March 22, 2018.

71111.05AQ - Fire Protection Annual/Quarterly

Quarterly Inspection (4 Samples)

The inspectors evaluated fire protection program implementation in the following selected

areas:

(1) Compartment 25-1, Units 1,2, and 3, 550' Intake Pumping Station and 565' Component

Cooling Water (CCW) Pump Deck on January 26, 2018

(2) Unit 2 RHR Heat Exchanger 2B, 2D, 2A, and 2C Rooms Elevation 565 and 593, and

Area 2-4 - South of Q - Unit 2 Elevation 593 on February 21, 2018

(3) Unit 2 Auxiliary Instrument room, Fire Area 16-M on March 14, 2018

(4) Compartment 26-A, Units 1, 2 and 3 Turbine Building on March 16, 2018

Annual Inspection (1 Sample)

(1) The inspectors evaluated fire brigade performance on March 6, 2018. The Browns Ferry

Fire brigade responded to report of smoke coming from a motor for the Unit 1/2 Diesel

Building CO2 tank compressor.

71111.06 - Flood Protection Measures

Internal Flooding (1 Sample)

(1) The inspectors evaluated internal flooding mitigation protections in the Unit 2 480V

Shutdown Board Rooms on February 2, 2018

Cables (1 Sample)

The inspectors evaluated cable submergence protection in:

(1) Hand holes 15 and 26 containing underground cables on January 8, 2018

71111.08 - Inservice Inspection Activities (1 Sample)

The inspectors evaluated boiling water reactor non-destructive testing by observing or

reviewing the following examinations from February 28 to March 1, 2018:

(1) Magnetic Particle Examination (MT)

a) MT of Weld HPCI-3-009-003 C1R2, Work Order (WO) 117544712, American

Society of Mechanical Engineers (ASME) Class 2. This review involved a pressure

boundary weld. (Reviewed)

(2) Liquid Penetrant Examination (PT)

a) PT of Weld RWCU-3-001-078 C1R0, WO 117656145 ASME Class 1. This review

involved a pressure boundary weld. (Reviewed)

(3) Radiographic Examination (RT)

a) RT of Weld HPCI-3-009-003 C1R0, WO 117544712, ASME Class 2. This review

involved a pressure boundary weld. (Reviewed)

b) RT of Weld HPCI-3-009-003 C1R1, WO 117544712, ASME Class 2. This review

involved a pressure boundary weld. (Reviewed)

c) RT of Weld HPCI-3-009-003 C1R3, WO 117544712, ASME Class 2. This review

involved a pressure boundary weld. (Reviewed)

(4) Ultrasonic Test (UT)

a) UT Examination Report R-049, Pipe to Elbow Weld, Component ID: DSRHR-3-04.

ASME Class 1. This review involved a pressure boundary weld. (Reviewed)

b) UT Examination Report R-085, Nozzle to Shell Weld, Component ID: N3D-NV.

ASME Class 1. This review involved a pressure boundary weld. (Observed)

(5) Visual Test (VT)

a) VT Examination Report R-033, Pipe Support, Component ID: 3-478400-099. ASME

Class 1. (Reviewed)

b) VT Examination Report R-069, Variable Spring Can, Component ID: RHR-3-H-146.

ASME Class 1. (Reviewed)

71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance

Operator Requalification (1 Sample)

The inspectors observed and evaluated a licensed operator requalification exam session for

the Group 0 operating crew on the Unit 3 Simulator involving a stuck open main steam relief

valve, inadvertent high pressure coolant injection actuation, unit board trip and Anticipated

Transient Without Scram (ATWS) with main steam isolation valves open on January 4,

2018.

Operator Performance (1 Sample)

The inspectors observed and evaluated startup of the Unit 3 reactor on January 12, 2018,

Unit 3 turbine control valve manipulations and power maneuvering on January 26, 2018,

shutdown of Unit 3 on February 17, 2018, and startup of the Unit 1 reactor on March 21,

2018.

71111.12 - Maintenance Effectiveness

Routine Maintenance Effectiveness (2 Samples)

The inspectors evaluated the effectiveness of routine maintenance activities associated

with the following equipment and/or safety significant functions:

(1) Unit 3 Turbine Stop and Control Valves. Maintenance Rule (MR) Function 047-B and

history of vibrations causing problems.

(2) MR Functions for System 575, 4kV Power Supply and Busses in (a)(1) status

71111.13 - Maintenance Risk Assessments and Emergent Work Control (9 Samples)

The inspectors evaluated the risk assessments for the following planned and emergent

work activities:

(1) Planned risk, associated with inoperable Main Bank Battery 3 and Battery Board 3 on

January 2, 2018

(2) Emergent work associated with oscillations of the Unit 3 number 3 Control Valve on

January 22, 2018

(3) In-office review of proposed Unit 3 refueling outage risk plan

(4) Shutdown risk associated with Unit 3 on shutdown cooling and reactor water level

control at 80 inches on February 17, 2018 (day 1) with Time to Boil at 37 minutes

(5) Reactor pressure vessel head lift on February 18, 2018

(6) Shutdown risk associated with Unit 3 during Operations with Potential for Draining the

Reactor Vessel (OPDRV) for replacing B Recirculation Pump seals on

February 22, 2018

(7) Shutdown risk associated with Unit 3 during OPDRV for replacing 32 control rod drives

on February 27, 2018

(8) Yellow shutdown risk during planned maintenance on Unit 3 Division II 4160V boards on

March 10, 2018

(9) Yellow shutdown risk on U-1 with short time to boil with Unit 3 still in a refueling outage

on March 20, 2018

71111.15 - Operability Determinations and Functionality Assessments (5 Samples)

The inspectors evaluated the following operability determinations and functionality

assessments:

(1) Unit 2, HPCI valve 73-44, design opening thrust exceeding the bearing rating and the

associated operator work around (OWA) on January 22, 2018

(2) Incorrect RHR system pressure gage used for verification of Technical Specification

surveillance test requirements (Condition Report (CR) 1372616, 1373852) on January 5,

2018

(3) Turbine Control Valve Fast Closure channel operability with Unit 3 turbine control valve

control circuit fuse and wiring changes (CR 1379519, 1382150) on January 22, 2018

(4) APRM 4 fault and 2-out-of-4 voter number 4 operability (CR 1382124) on January 30,

2018

(5) Past operability evaluation for diesel generator 3C load acceptance test failure (CR

1389131)

71111.18 - Plant Modifications (1 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) DCN 69424 Replace Unit 3 Condenser Vacuum Pressure Switches with Pressure

Transmitters

71111.19 - Post Maintenance Testing (10 Samples)

The inspectors evaluated the following post maintenance tests:

(1) Unit 3 4kV Shutdown Board 3EB loss of power logic system test on March 6, 2018

(2) Testing of Unit 3 overhauled motor operated valve 74-53, RHR Loop I Low Pressure

Coolant Injection Valve

(3) Testing of replacement Unit 3 Division I Emergency Core Cooling System (ECCS)

Inverter

(4) Local leak rate test of 3-FCV-73-45 HPCI discharge check valve following installation of

softer seat material.

(5) Unit 3 Emergency High Pressure Make-Up Basic Pump Recirculation Testing

(6) Surveillance 3-SR-3.1.7.3 Standby Liquid Control System Enriched Sodium Pentaborate

Solution Concentration, Quantity Calculation, and ATWS Equivalency Calculation

following implementation of modified boron enrichment for Extended Power Uprate

(7) Test of DCN to install parallel auxiliary contact for 3EC 4kv shutdown board normal

feeder breaker 1338

(8) Testing of Unit 3 overhauled motor operated valve 74-73, RHR Loop II Test Outboard

Isolation Valve

(9) Testing of Unit 3 overhauled motor operated valve 73-2, HPCI Turbine Steam Supply

Inboard Primary Containment Isolation Valve

(10) Testing of replacement STA switch on 3EC diesel generator output breaker

71111.20 - Refueling and Other Outage Activities (Partial Sample)

The inspectors evaluated refueling outage U3R18 activities from February 16, 2018

through March 31, 2018. The inspectors completed inspection procedure sections

03.01.a, b, c, d and e.2.

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Routine (2 Samples)

(1) 3-SR-3.8.1.9, (3B OL) Unit 3 EDG load acceptance test, on February 6, 2018,

(2) 3-SR-3.3.1.1.13 APRM 1-4 calibrations per DCN 68463 Stage 4 associated with the

Extended Power Uprate (EPU) modification on March 13, 2018

In-service (3 Samples)

(1) 0-SI-4.5.C.1(A2-COMP) - Residual Heat Removal Service Water (RHRSW) Pump A2

IST Comprehensive Pump on January 2, 2018

(2) 1-SR-3.5.1.6 (RHR II) - Quarterly RHR System Rated Flow Test Loop II, on February 7,

2018

(3) 3-SR-3.1.7.7, Unit 3 Standby Liquid Control system functional test on March 22, 2018

RADIATION SAFETY

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (1 Sample)

The inspectors evaluated radiological hazards assessments and controls.

Instructions to Workers (1 Sample)

The inspectors evaluated worker instructions.

Contamination and Radioactive Material Control (1 Sample)

The inspectors evaluated contamination and radioactive material controls.

Radiological Hazards Control and Work Coverage (1 Sample)

The inspectors evaluated radiological hazards control and work coverage.

High Radiation Area and Very High Radiation Area Controls (1 Sample)

The inspectors evaluated risk-significant high radiation area and very high radiation area

controls.

Radiation Worker Performance and Radiation Protection Technician Proficiency (1 Sample)

The inspectors evaluated radiation worker performance and radiation protection technician

proficiency.

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

and Transportation

Radioactive Material Storage (1 Sample)

The inspectors evaluated the licensees radioactive material storage.

Radioactive Waste System Walk-down (1 Sample)

The inspectors evaluated the licensees radioactive waste processing facility during plant

walkdowns.

Waste Characterization and Classification (1 Sample)

The inspectors evaluated the licensees radioactive waste characterization and

classification.

Shipment Preparations (1 Sample)

The inspectors evaluated the licensees radioactive material shipment preparation

processes.

Shipment Records (1 Sample)

The inspectors evaluated the licensees non-excepted package shipment records.

OTHER ACTIVITIES - BASELINE

71151 - Performance Indicator Verification

The Resident Inspectors verified licensee performance indicators submittals listed below

for the period from January 1, 2017, through December 31, 2017. (6 Samples)

(1) Units 1, 2, and 3 Reactor Coolant System Leakage

(2) Units 1, 2, and 3 Reactor Coolant System Activity

The inspectors reviewed licensee PI submittals listed below for the period from

April 1, 2017, through February 12, 2018. (1 Sample)

(1) OR01: Occupational Exposure Control Effectiveness

71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues (5 Samples)

The inspectors reviewed the licensees implementation of its corrective action program

related to the following issues:

(1) Unresolved Item (URI) 05000260, 296/2017008-01, Potential Inadequate Configuration

Control of the Unit 2 and Unit 3 HPCI Discharge Valves

(2) URI 05000260, 296/2017008-02, Potential Inadequate Commercial Grade Dedication of

Components in Safety Related Valves

(3) URI 05000260, 296/2017008-03, Potential Inadequate Configuration Control of the Unit

and Unit 3 HPCI Discharge Valves

(4) URI 05000296/2017008-04, Potential Inadequate Operator Response to Inadvertent

HPCI Injection

(5) Problem Identification & Resolution and Regulatory Commitments associated with Unit 3

Extended Power Uprate

71153 - Follow-up of Events and Notices of Enforcement Discretion

Events (3 Samples)

(1) The inspectors evaluated the plant response and licensees response for a Unit 3 reactor

scram on January 10, 2018.

(2) The inspectors responded to a Notice of an Unusual Event after a routine search of a

work-related vehicle noted a suspicious object underneath the vehicle. It was later

determined the suspicious object was a normal part of the vehicle

(3) The inspectors evaluated the plant response and licensees response for a Unit 1 reactor

scram on March 18, 2018.

Licensee Event Reports (2 Samples)

The inspectors evaluated the following licensee event reports (LER) which can be accessed

at https://lersearch.inl.gov/LERSearchCriteria.aspx:

(1) LER 05000260/2017-002-00, Inoperable Primary Containment Isolation Valve Resulting

in Condition Prohibited by Technical Specifications

(2) LER 05000259/2016-004-01, Incorrect Tap Settings for 480 Volt Shutdown Transformer

Results in Inoperability of Associated 480V Shutdown Boards

OTHER ACTIVITIES - TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

71004 - Power Uprate

Erosion-Corrosion/Flow-Accelerated-Corrosion Monitoring Programs (2 samples)

Inspectors reviewed the Erosion Corrosion/Flow-Accelerated Corrosion (EC/FAC) program

in accordance with the guidance contained in NRC Inspection Procedure 49001, Inspection

of Erosion-Corrosion/Flow-Accelerated-Corrosion Monitoring Programs dated 12/11/98.

Summary of Power Uprate Inspection Samples Contained in this Report:

Integrated Plant Operations at the Uprated Power Level (Unit 3) (1 sample)

(1) Licensed Operator Requalification Training for EPU (Section 71111.11)

Plant Modifications (all Units) (1 sample)

(1) DCN 69424 Replace Unit 3 Condenser Vacuum Pressure Switches with Pressure

Transmitters (Section 71111.18)

Post-Maintenance / Post-Modification or Surveillance Tests (Unit 3) (2 samples)

(1) Surveillance 3-SR-3.1.7.3 Standby Liquid Control System Enriched Sodium Pentaborate

Solution Concentration, Quantity Calculation, and ATWS Equivalency Calculation

following implementation of modified boron enrichment for Extended Power Uprate

(Section 71111.19)

(2) 3-SR-3.3.1.1.13 APRM 1-4 calibrations in accordance with DCN 68463 Stage 4

associated with the EPU modification (Section 71111.22)

Regulatory Commitments and Recommended Areas for Inspection (all Units) (1 sample)

(1) Regulatory Commitments related to EPU (Section 71152)

Identification and Resolution of Problems (Unit 3) (1 sample)

(1) Problem Identification and Resolution related to EPU (Section 71152)

Flow Accelerated Corrosion and Erosion Corrosion Program Reviews (all Units) (2 samples)

(1) Flow Acceleration Corrosion Program (Section 71004)

(2) Erosion Corrosion Program (Section 71004)

INSPECTION RESULTS

71111.19 - Post Maintenance Testing

Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches

Cornerstone Significance Cross-cutting Report

Aspect Section

Mitigating Green [H.1] - 71111.19

Systems NCV 05000259, 260, 296/2018001-01 Resources

Closed

Introduction: A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion V, was

identified when the licensee failed to perform an adequate post-maintenance test in

accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post

maintenance testing on the 3C diesel generator output breaker did not ensure that all

contacts on replacement stationary switches successfully changed state after installation.

Description: On February 20, 2018, during the biannual performance of TS SR 3.8.1.9 for the

3C diesel generator, several automatic safety functions did not occur as designed. The 3B

RHR, 3B Core Spray, and B1 RHRSW pumps did not automatically start after the 3C diesel

generator output breaker closed in to the 3EC 4kV Shutdown Board. The Unit 3 480V Load

Shed for Division II also did not occur. The degraded condition was determined to be the

result of one pair of contacts on the diesel generator output breakers stationary switch failing

to make up when the breaker closed in. Troubleshooting revealed that the stationary switch

contact failed to make up because the associated actuating arm on the breaker failed to

rotate the stationary switch sufficiently. Although these actuations did not automatically

occur, they could have been accomplished manually once recognized by control room

operators.

This particular contact was used in a part of the logic circuitry to signify that the diesel

generator had successfully tied onto the board and was ready to accept the designed safety

loads when there was an accident signal present and normal offsite power to the board was

not available. The contact also initiates load shedding of non-essential 480 volt loads to

prevent the diesel generator from being overloaded as the safety loads are automatically

sequenced on. Additionally, because the 3B Core Spray pump would not have automatically

started, the 3D Core Spray pump would also not have automatically started because of the

design of the Core Spray initiation logic. The last time that the switch was known to be

working correctly was during the last biannual surveillance test in February of 2016. The

licensees past operability evaluation concluded that the 3C diesel generator, 3B and 3D Core

Spray pumps, 3B RHR pump, B1 RHRSW pump, and Unit 3 480V Division II Load Shed

Logic be considered inoperable from February 25, 2016, until February 20, 2018.

From a review of historical maintenance on this breaker, it was identified that the switch was

replaced on March 3, 2016, via work order 116872223 as a 24 year preventative maintenance

action; however, only a portion of the switchs contacts were tested for continuity during the

post-maintenance tests. Inspectors identified that the testing performed did not satisfy the

requirements of NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, section

3.2.2.A.5 required that, PMTs for safety-related circuits shall include testing to ensure

affected portions of the logic circuitry are tested if they were potentially affected.

Corrective Action(s): The breaker stationary switch was replaced and retested satisfactorily.

Corrective Action Reference(s): CR 1389131

Performance Assessment:

Performance Deficiency: The failure to perform adequate post maintenance testing on the 3C

diesel generator output breaker in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance

Testing, was a performance deficiency.

Screening: The performance deficiency was more than minor because it was associated with

the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected

the cornerstone objective of ensuring availability, reliability and capability of systems that

respond to initiating events to prevent undesirable consequences. Specifically, the

performance deficiency caused the licensee to return a safety-related breaker to service that

was later discovered to not be able to perform all of its safety related functions and rendered

multiple supported components inoperable.

Significance: Using Chapter 0609, Appendix A, The Significance Determination Process for

Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the issue screened

as requiring a detailed risk evaluation because it resulted in an actual loss of function of at

least a single train for greater than its TS allowed outage time. An NRC Regional Senior

Reactor Analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6

and SPAR Model Version 8.50 for Unit 3. The SRA modeled the condition by assuming the

EDG 3C Load Sequencer was failed for one year, which accounted for pump automatic start

failures, and that manual start remained available. To account for potential manual start

failures, the SRA performed a human reliability analysis using the SPAR-H method and

adjusted the model to include a probability of operator failure to recover the sequencer. The

dominant sequences (12), which accounted for 90% of the change, involved loss of offsite

power with failure of various EDG combinations leading to a station blackout, loss of

suppression pool cooling, and failure of low pressure injection. The result was a change in

core damage frequency of less than 1E-7/year and was primarily mitigated by operator

recovery. Because the change was less than 1E-7/year, no further analysis was needed for

external events or large early release, and this finding was determined to be of very low

safety significance (Green).

Cross Cutting Aspect: [H.1] - Resources. The apparent cause of the performance deficiency

was that leaders did not ensure that plant procedures contained guidance for developing

adequate post-maintenance tests for breaker stationary switch replacements.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions Procedures and

Drawings, states, in part, that instructions shall include appropriate quantitative and qualitative

acceptance criteria for determining that important activities have been satisfactorily

accomplished. Contrary to the above, on March 3, 2016, work order 116872223 did not

contain post-maintenance test instructions with appropriate acceptance criteria for

determining that the breaker stationary switch replacement had been satisfactorily

accomplished.

Enforcement Actions: This violation is being treated as a non-cited violation, consistent with

Section 2.3.2 of the Enforcement Policy.

71124.01 - Radiological Hazard Assessment and Exposure Controls

Unauthorized Entry into a High Radiation Area (HRA)

Cornerstone Significance Cross-cutting Report

Aspect Section

Occupational Green [H.8] - 71124.01

Radiation Safety NCV 05000259, 260, 296/2018001-02 Procedure

Closed Adherence

Introduction: A self-revealing, Green, NCV of Technical Specification (TS) 5.7.1, was

identified for a worker who entered a HRA without proper authorization. Specifically, the

worker entered the Unit 3 A & C Residual Heat Removal (RHR) heat exchanger room using

an incorrect Radiation Work Permit (RWP) and without being briefed on the radiological

conditions.

Description: On March 24, 2018, an electrician was assigned the job of installing a jumper on

a component in the Unit 3 A & C RHR heat exchanger room. At the time, this area was

posted Contaminated Area and High Radiation Area. The electrician logged into RWP

18370011, which did not allow entry into HRAs. The worker also bypassed the Radiation

Protection (RP) desk and failed to receive a briefing on radiological conditions in the area.

The worker then dressed in anti-contamination clothing and proceeded past the HRA

boundary into the room. He subsequently received a dose rate alarm of 82 mrem/hr, which

exceeded the ED alarm setpoint of 60 mrem/hr, and immediately exited the area. A RP

technician performed a follow up survey and confirmed the presence of HRA dose rates up to

300 mrem/hr at 30 cm.

Corrective Action(s): The licensee took immediate corrective actions including Radiologically

Controlled Area (RCA) access restriction for the individual and initiation of an investigation of

the event including surveys of the areas entered.

Corrective Action Reference(s): CR 1390579

Performance Assessment:

Performance Deficiency: The workers entry into a HRA without using an appropriate RWP

and without being briefed on radiological conditions in the area, as required by TS 5.7.1, was

a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the Occupational Radiation Safety Cornerstone attribute of

Human Performance and adversely affected the cornerstone objective of ensuring adequate

protection of worker health and safety from exposure to radiation from radioactive material

during routine civilian nuclear reactor operation.

Significance: The inspectors assessed the significance of the finding using Inspection

Manual Chapter (IMC) 0609 C, Occupational Radiation Safety Significance Determination

Process. The finding was not related to As Low As Reasonably Achievable (ALARA)

planning, nor did it involve an overexposure or substantial potential for overexposure, and the

ability to assess dose was not compromised. Therefore, the inspectors determined the finding

to be of very low safety significance (Green).

Cross-cutting Aspect: This finding involved the cross-cutting aspect of Human Performance,

Procedural Adherence, because the event was a direct result of the workers failure to adhere

to administrative requirements for HRA access.[H.8]

Enforcement:

Violation: Technical Specification 5.7.1 requires that access to HRAs be controlled by means

of an RWP and entry into such areas shall be made only after dose rates in the area have

been determined and entry personnel are knowledgeable of them. Contrary to this, on

February 24, 2018, a licensee employee entered a posted high radiation area without proper

RWP authorization and without being knowledgeable of the radiological conditions. Upon

identification, the licensee immediately implemented RCA access restrictions for the

individual and completed follow up surveys of the areas entered.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2

of the Enforcement Policy.

Failure to Implement Controls for Locked High Radiation Area (LHRA) Access

Cornerstone Significance Cross-cutting Report

Aspect Section

Occupational Green [H.4] - 71124.01

Radiation Safety NCV 05000259/260/296/ 2018001-03 Teamwork

Closed

Introduction: A self-revealing, Green, NCV of TS 5.7.2, was identified for the failure to control

access to a LHR

A. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell

without RP personnel present. In doing so, the worker accessed an area with dose rates >1

rem/hr that had not been posted, locked, or surveyed prior to entry.

Description: On February 18, 2018, a carpenter was directed by the RP Drywell Coordinator

to install a ladder on the 563 elevation of the Unit 3 drywell near the A blower bank. The

inspectors noted the ladder allowed access to an area that had not been surveyed by RP,

was not posted or controlled as a LHRA, and no RP technician was present during the

installation. While climbing up the ladder to complete a tie off, the carpenter received a dose

rate alarm of 458 mrem/hr which exceeded the ED alarm setpoint of 400 mrem/hr. The ED

alarm was seen by the remote monitoring station and a roving RP technician was dispatched

to respond. The RP technician directed the carpenter to exit the drywell and report to RP.

The technician immediately performed a survey of the area accessible by the ladder and

discovered dose rates up to 20 rem/hr on contact and 6 rem/hr at 30cm.

Corrective Action(s): The licensee took immediate corrective actions including posting a

LHRA guard until appropriate controls could be implemented.

Corrective Action Reference(s): CR 1388425

Performance Assessment:

Performance Deficiency: The failure to post, lock, and survey the area prior to entry (or be

escorted by RP), as required by TS 5.7.2, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor

because it was associated with the Occupational Radiation Safety Cornerstone attribute of

Human Performance and adversely affected the cornerstone objective of ensuring adequate

protection of worker health and safety from exposure to radiation from radioactive material

during routine civilian nuclear reactor operation.

Significance: The inspectors assessed the significance of the finding using IMC 0609 C,

Occupational Radiation Safety Significance Determination Process. The finding was not

related to ALARA planning, nor did it involve an overexposure or substantial potential for

overexposure (due to the use of remote monitoring), and the ability to assess dose was not

compromised. Therefore, the inspectors determined the finding to be of very low safety

significance (Green).

Cross-cutting Aspect: This finding involved the cross-cutting aspect of Human Performance,

Teamwork, because the event was a direct result of poor coordination between work groups.

[H.4]

Enforcement:

Violation: Technical Specification 5.7.2 requires that HRAs with dose rates > 1 rem/hr at 30

cm, but less than 500 rad/hr at 1 m, be conspicuously posted and provided with a locked or

continuously guarded door. TS 5.7.2 also requires that, except for personnel escorted by RP,

entry into such areas be made only after dose rates in the area have been determined and

entry personnel are knowledgeable of them. Contrary to this, on February 18, 2018, a

licensee employee installed a ladder that allowed access to an area with dose rates > 1

rem/hr at 30 cm, but less than 500 rad/hr at 1 m, that was not posted or locked. In addition,

the employee entered the area without a RP escort and prior to dose rates being determined.

The licensee took immediate corrective actions including posting a LHRA guard until

appropriate controls could be implemented.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2

of the Enforcement Policy.

71152 - Problem Identification and Resolution

Inadequate Configuration Control of HPCI Valve Design Issues

Cornerstone Significance Cross-cutting Report Section

Aspect

Mitigating Green None 71152 - Annual

Systems NCV 05000296/2018001-04 Follow-up of

Closed Selected Issues

Introduction: A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, was

identified when the licensee failed to ensure adequate control of valve design configurations

in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control

Revision 6. Specifically, the licensee machined a HPCI discharge valve contrary to original

design and did not document the change.

Description: On September 24, 2017, during the performance of the quarterly HPCI pump

test an unintentional injection of colder condensate water into the reactor vessel occurred

causing reactor power to be at 104% power for about 5 minutes. The injection was caused by

a fractured actuator yoke nut that had developed during the June 2017 stroke test of 3-FCV-

73-44 leaving the valve partially open. The licensee disassembled and inspected 3-FCV-73-

44, and three other valves as a part of their extent of condition review.

During the disassembly of the valves, the licensee identified that the yoke nut flanges on two

of the valves were found to be 1 versus that specified in the original vendor drawing which

showed the flange was 1.25. The licensees evaluation determined that during past

modifications of these valves the yoke nuts were received from the vendor and machined

down to 1 without approval or documentation. Licensee extent of condition reviews identified

another HPCI valve with an unapproved and undocumented 0.25 spacer below the bottom

bearing set. Other deviations identified, were missing ball bearings and additional

components in the bearing housing (bearing cage).

Corrective Action(s): As an immediate corrective action the licensee restored each of the

valves to their original configurations in accordance with the vendor drawings.

Corrective Action Reference(s): CRs 1341458, 1357076, 1347334, and 1359556

Performance Assessment:

Performance Deficiency: The failure to ensure adequate control of valve design

configurations, as required by NPG-SPP-9.3 revision 6, was a performance deficiency.

Specifically, the licensee machined a HPCI discharge valve contrary to original design and did

not document the change.

Screening: The performance deficiency was more than minor because it was associated with

the design control attribute and affected the associated cornerstone objective to ensure

availability, reliability and capability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, the resulting yoke nut and bearing configuration

contributed to the failure of the valve, and prevented the valve from stroking fully closed.

Significance: Using Chapter 0609, Appendix A, The Significance Determination Process for

Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the issue screened

as having very low safety significance (Green) because it was a deficiency that affected the

design and qualification of safety related, HPCI valves, but operability was maintained.

Cross Cutting Aspect: No cross cutting aspect was assigned to this finding because the

inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that

measures shall include provisions to assure that appropriate quality standards are specified

and included in design documents and that deviations from such standards are controlled.

NPG-SPP-9.3 establishes a process of administrative controls and regulatory/quality

requirements for plant modifications and changes to engineering documents. NPG-SPP-9.3

Rev. 6, Step 3.1.9.A.1 states, in part, that vendor manuals and configuration control design

documents affected by the change package have been revised or updated. Contrary to the

above, in April 2012, the licensee failed to ensure that vendor manuals and other

configuration control design documents affected by the change were revised or updated for 3-

FCV-73-44.

Enforcement Actions: This violation is being treated as an NCV, consistent with Section 2.3.2

of the Enforcement Policy.

This finding closes URI 05000260, 296/2017008-03

Unresolved Item URI 05000260, 296/2017008-01, Potential 71152 - Annual

(Closed) Inadequate Weak Link Analysis for Unit 2 and Unit Follow-up of

3, HPCI Discharge Valves Selected Issues

Description: The subject URI was identified to determine if a performance deficiency exists

regarding the adequacy of the weak link analysis for the valve and actuator of the HPCI Unit 2

and Unit 3 discharge valves. Inspectors reviewed the various historical weak link analyses for

these valves. The original vendor analysis only included the results for the most limiting part

in the valve rather than a complete documented analysis for each area analyzed. This

minimal level of documentation met the licensees and regulatory standards. As a result, the

licensee had no documentation that would cause engineers to believe that the valves yoke

nut or yoke nut bearings would exceed their load ratings once the valves actuator thrust was

increased in 2012. The valve vendor failed to recognize these loading limitations during their

reviews that supported the licensees thrust modification. As a result of this discovery, Crane

Nuclear Inc. issued a 10 CFR Part 21 Notification of Defect to the NRC on December 19,

2017.

Corrective Action Reference(s): CR 1344131

Closure Basis: Inspectors concluded that the defects described in the valve vendors

notification were not reasonably within the licensees ability to foresee and did not represent a

performance deficiency.

Unresolved Item URI 05000296/2017008-02, Potential Inadequate 71152 - Annual

(Closed) Commercial Grade Dedication of Components in Follow-up of

Safety Related Valves Selected Issues

Description: The subject URI was identified to determine if a performance deficiency existed

regarding the adequacy of the commercial grade dedication of the valve yoke nut bearings in

the HPCI discharge valves on Unit 2 and Unit 3.

Corrective Action Reference(s): CR 1358257

Closure Basis: Since the original thrust bearings were purchased/provided directly from the

valve manufacturer, the licensees commercial grade dedication process was not applicable

and there was no performance deficiency attributable to the licensee associated with the

variation in bearing configuration. The acceptability of the valve manufacturers dedication

process for the commercial grade bearings was not within the scope of this inspection.

Replacement bearings were procured after the as-found configurations were discovered to be

different than the original design configuration. These replacement bearings were procured

as commercial grade items and dedicated by the licensee prior to installation. No findings

were identified.

Unresolved Item URI 05000296/2017008-04, Potential Inadequate 71152 - Annual

(Closed) Operator Response to Inadvertent HPCI Injection Follow-up of

Selected Issues

Description: The subject URI was identified to determine if a performance deficiency exists

regarding the adequacy of control room operators response to the September 24, 2017, Unit

inadvertent HPCI system injection into the reactor vessel.

Prior to the surveillance, reactor power had been reduced to 99.3 percent. The inadvertent

injection caused reactor power to exceed the 100 percent licensed thermal power limit (RTP)

and initiated an alarm for a reactor feedwater control system input failure. After the alarm,

operators noticed that the HPCI check valve 3-73-45 was indicating open despite the

upstream discharge valve 3-FCV-73-44 indicating closed. Once the operators diagnosed that

HPCI injection was occurring, they initiated a HPCI turbine trip. The HPCI injection lasted

approximately five minutes and reactor power stabilized at 104.8 percent. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

average RTP was less than 100%.

The inspectors reviewed the licensees performance analysis, Regulatory Information

Summary (RIS) 2007-21, Adherence to Licensed Power Limits and IMC 0612, Appendix E,

Examples of Minor Issues which discussed this circumstance. The training analysis concluded

that the crew did not understand the expected plant response with a HPCI injection and thus

were delayed in performing actions specified in AOI-3-1, Loss of Reactor Feedwater. Step 15

directed tripping the HPCI pump. The RIS stated that thermal power may rise slightly due to

normal changes in plant parameters and operators are expected to take prompt corrective

action to reduce thermal power once it is discovered to be above the licensed limit. Licensees

may not intentionally operate or authorize operation above the maximum power level as

specified in the license.

IMC0612, Appendix E found this circumstance to be one of minor significance when:

• Operators had performed the prerequisite power reduction and after realizing that

thermal power had exceeded RTP, promptly decreased thermal power below the RTP.

• Operators made appropriate and timely adjustments to prevent the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average CTP

from exceeding RTP

Corrective Action Reference(s): CR 1346991

Closure Basis: The Inspectors concluded that there was no intentional operation above RTP

and that the operator response met the guidance in both the RIS 2017-21 and the IMC 0612,

Appendix E.

Observation 71152 - Annual Follow-up of Selected Issues

For the implementation of Unit 3 extended power uprate, inspectors assessed the licensees

performance regarding problem identification and resolution against selected attributes listed

in section 03.06 of Inspection Procedure 71152. Inspectors reviewed condition reports

associated with extended power uprate to verify that problems were being promptly identified,

evaluated, prioritized and resolved within the licensees corrective action program. Inspectors

also reviewed the NRC Safety Evaluation for any regulatory commitments associated with

extended power uprate and found that the licensee did not make any regulatory commitments.

Overall, inspectors found no licensee performance weaknesses during this review.

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public

disclosure.

• On January 25, 2018, the EC/FAC inspection results were presented to Steve Bono and

other members of the licensee staff

• On March 2, 2018, the radiation protection inspection and in-service inspection results were

presented to Mr. D. L. Hughes and other members of the licensee staff.

• On April 20, 2018, the quarterly resident inspector inspection results were presented to Mr.

Werner Paulhardt and other members of the licensee staff.

DOCUMENTS REVIEWED

IP 71111.04

Procedures

3-OI-74, Residual Heat Removal System, Revision 125

0-OI-72, Auxiliary Decay Heat Removal System, Revision 60

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

Drawings

3-47E811-1, Flow Diagram Residual Heat Removal System, Revision 72

0-47E873-1, Flow Diagram Aux Heat Removal System, Revision 8

0-15E740-1, Single-Line Diagram ADHR Service Entrance and MCC, Revision 13

Other Documents

CR 1388305

0-BFN-VM-5003, Installation, Operation and Maintenance Instructions and Engineering

Document Package

FSAR Chapter 8.4, Normal Auxiliary Power System

DCN 71673, Implementation of U3 Emergency High Pressure Make-Up Pump System

IP 71111.05

Procedures

Browns Ferry Fire Protection Report-VOLUME 2, Fire Protection Report Volume 2, Revision 58

Other Documents

MDN0009992012000100, Browns Ferry Nuclear Power Plant, Units 1, 2, and 3, Fire Risk

Evaluations, Revision 6

EDQ099920110010, NFPA 805 - Nuclear Safety Capability Analysis, Revision 33

IP 71111.06

Drawings

2-47W2392-642, Fire Protection - 10CFR50 Appendix R Penetration Seal Tabular Drawings

E

L. 621.25, Revision 2

0-47W510-1, Mechanical Roof Drains, Revision 1

0-47W510-2, Mechanical Roof Drains, Revision 4

Other Documents

BFN-57250, BFN-0-PMP-040-0031, Visual Inspection of Listed Handholes and Sumps Per

95003 Commitment, Revision 6

WO 118861289

CR 1375311

CR 1375316

NDN-000-999-2007-0031, Internal Flooding BFN Probabilistic Risk Assessment, Revision 0

DED-TM-PF2, Concluding Report of the Effects of Postulated Pipe Failure Outside of

Containment for the Browns Ferry Nuclear Plant Unit s 2 and 3, dated March 1, 1974

IP 71111.08

Procedures

N-UT-64, Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds, Revision

0016

N-UT-78, PDI Generic Procedure for the Manual Ultrasonic Examination of Reactor Pressure

Vessel Welds, PDI-UT-6, Revision 9

N-UT-90, Generic Procedure for the Ultrasonic Detection and Sizing of Reactor Pressure Vessel

Nozzle to Shell Welds and Nozzle Inner Radius, Revision 003

N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Revision

0047

PDI-UT-2, PDI Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds,

Revision H, October 4, 2017

PDI-UT-6, PDI Generic Procedure for the Ultrasonic Examination of Reactor Pressure Vessel

Welds, Revision I, August 1, 2017 PDI-UT-11, Generic Procedure for the Ultrasonic

Examination of Reactor Pressure Vessel Nozzle-to-Shell Welds and the Nozzle Inner Corner

Radius, Revision D 08-01-2017, Revision D, August 1, 2017

Drawings

2-47W2392-6

Other Documents

CDQ0-003-2006-0002, BFN RPV Feedwater Nozzles Fatigue and Fracture Mechanics

Evaluation, Revision 14

CR 1135166, ISI Examination Drawings

CR 1143845, CR to Track Accept-As-Is for Indication on Top of Unit 3 Vessel Head

CR 1145011, FME Voluntary Stop Work for 3A Feed Water Heater Welding

CR 1145022, FME Procedure Not Followed by Contractors

CR 1145738, Incorrect Detail Weld Procedure Specification

CR 1146291, Documentation Errors on Weld Data Sheets

CR 1146995, Tack Welds Made without Sufficient Purging

CR 1146888, Potential Rework Event

CR 1147745, Discrepancies and Errors on Weld Data Sheet

CR 1147756, A D&Z Mods Welder Contaminated in RCA Clean Area

CR 1148490, U3R17 Jet Pump Wedge Wear and Set Screw Gaps / Indications

CR 1150215, Welding Being Performed without a Fire Watch

CR 1150705, NOI U3R17-007: Moisture Seal Barrier (MSB) Loss of Adhesion.

CR 1166944, Core Shroud Off-Axis Cracking Interim Inspection & Flaw Evaluation Guidance

CR 1184618, Through-Wall Penetration in Safety-Related Heat Exchanger Shell

CR 1187114, Part 21 - Inadequate Vendor Documentation of Far Vision Acuity Certifications

CR 1210910, Potential Code Class-2 Piping Leak on RBCW Piping @ 1-DRV-70-507

Connection Elbow

CR 1221309, Two Welding Machines Left On and Unattended

CR 1223258, Invertec V350 Pro Welder Left On and Unattended

CR 1227532, Scheduled Containment ISI Examination Not Performed

CR 1229969, Leakage Coming from 1-CKV-73-45

CR 1244822, Welding Sparks Escaped Containment Tent on RFF

CR 1250683, Request for Review of BWRVIP Position Regarding Aging Management of

Orificed Fuel Support Castings

CR 1284288, Re-Welding Stainless Steel Multiple Times Presents Various Issues

CR 1324316, (CRP-ENG-FSA-17-004) ISI Program Deficiencies

CR 1326645, BWRVIP Skip Outage Project Initiation

CR 1333664, BFN Leak Source Evaluation

Browns Ferry Nuclear Standard ISI Plan (Baseline) Standard Code ASME Section XI, 2007 Ed /

2008 Add Category Scheduling Compliance

Calibration Block WB-084 As-Built Verification Documentation

Certification for Magnaflux Ultragel II, Batch Number 16H031

Certificate of Compliance for Miniature Angle Beam Block, Serial Number 789631

Certificate of Compliance for Miniature Angle Beam Block, Serial Number 791719

Certificate of Conformity I07120001 for Visual Illumination Cards

Certified Material Test Reports for weld rods used for WOs 117544712 and 117656145

CRP-ENG-FSA-17-004, Focused Self-Assessment Report, Inservice Inspection at Browns

Ferry, Approved September 14, 2017

Detail Welding Procedure Specification (DWPS) GT88-O-1-N, Manual Gas Tungsten Arc

Welding, Revision 5

Drawing 3-47B400-99, Mechanical, Main Steam System Pipe Support, Revision 000

Drawing BF-18, Calibration Blocks As-Builts BF-18, Material: A-533, Revision 01

IVVI Examination Checklist Browns Ferry Unit 3 R18 Spring 2018 (BF3R18) Outage

Krautkramer Transducer Certification for Number 01FH9V

Krautkramer Transducer Certification for Number 16B003AA

Krautkramer Transducer Certification for Number 16B003AC

Krautkramer Transducer Certification for Number 16B003AG

Letter to TVA from NRC, dated March 14, 2017, Subject: Browns Ferry Nuclear Plant, Units 2

and 3 - Request for ASME Code,Section XI, Alternatives 2-ISl-30 and 3-ISl-27 for the Periods

of Extended Operation Regarding Reactor Pressure Vessel Circumferential Shell Weld

Examinations

Owners Activity Report for BFN, Unit 3, Cycle 17 Operation, dated 6/21/16

NDE Personnel Qualifications for

J. Hoover,

M. Kleinjan,

D. Maclean, D. Sawatzky

Report of Calibration for Krautkramer Ultrasonic Flaw Detector, Serial Number 0100H4

Report of Calibration for Krautkramer Ultrasonic Flaw Detector, Serial Number 0132M6

Report of Calibration for TEGAM Digital Thermometer, Serial Number T-257196

Report of Calibration for Keithley Digital Thermometer, Serial Number T-12463

UT Examination Report R-049, Pipe to Elbow Weld, Component ID: DSRHR-3-04

UT Examination Report R-085, Nozzle to Shell Weld, Component ID: N3D-NV

VT Examination Report R-033, Pipe Support, Component ID: 3-478400-099

VT Examination Report R-069, Variable Spring Can, Component ID: RHR-3-H-1

Welder Qualification Records for

C. Brock,

K. Davenport,

J. Gautney,

C. Hill,

C. Lindsey,

J. Parker,

S. Laird, and E. Woods

Weld Map and Data Sheets for WOs 117544712 and 117656145

Welding Procedure Qualification Record GTA 88-0-1, Gas Tungsten Arc Welding, dated

December 29, 1978

Welding Procedure Qualification Record GTA 88-0-5, Gas Tungsten Arc Welding, dated

April 15, 2004

WO 117544712, HPCI Mod per DCN 71865, Valve 73-23 and 73-603 to be Relocated

WO 117656145, Replace Valve BFN-3-TV-069-0583

IP 71111.11

Procedures

3-AOI-3-1, Loss of Reactor Feedwater or Reactor Water Level High/Low, Revision 12

3-AOI-1-1, Relief Valve Stuck Open, Revision 14

NPG-SPP-17.8.4, Conduct of Simulator Operations, Revision 4

BFN-ODM-4.20, Strategies for Successful Transient Mitigation, Revision 4

3-GOI-100-1A, Unit Startup, Revision 116

0-TI-248, Station Reactor Engineer, Revision 113

NPG-SPP-10.4, Reactivity Management Program, Revision 6

3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in

Power During Power Operations, Revision 61

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

3-OI-47, Turbine-Generator System, Revision 11

1-GOI-100-1A, Unit Startup, Revision 48

Other Documents

OPL175S055, SRV Fails Open, HPCI inadvertent actuation, 3B 4kV Unit Board Trip, ATWS with

MSIVs Open, Revision 0

IP 71111.12

Procedures

0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -

10CFR50.65, Revision 50

Other Documents

System Health Report for System 575 4kV AC Power Distribution,

U1/2&3 Function 575-B, C & E 4kV Power Supply Busses Sys (a)(1) Plan, Revision 11,

Effective October 27, 2017

Functional failure and Unavailability data for System 575 through February 2018

IP 71111.13

Procedures:

BFN-ODM-4.18 Protected Equipment, Revision 17

NPG-SPP-09.11.1 Equipment Out of Service Management, Revision 12

0-TI-248, Reactor Engineer, Revision 113

3-OI-47, Turbine-Generator System, Revision 111

MSI-0-000-LFT001, Lifting instructions for the control of heavy loads, Revision 0074

FSAR Appendix C, Structural Qualifications of Subsystems and Components, C.8, Control of

Heavy Loads

1-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Revision 8

3-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Revision 0017

NPG-SPP-10.6, Infrequently Performed Test or Evolutions, Revision 1

MCI-0-085-CRD001, Control Rod Drive Removal and Installation, Revision 0061

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

Drawings:

3-47E610-47-1, Mechanical Control Diagram Turbine, Revision 6

3-47E610-47-4, Mechanical Control Diagram Turbine, Revision 21

3-9952-582, Diagram Main Turbine Control Wiring, Revision A

Other Documents:

CR 1292238

Operator logs from May 4, 2017 through May 5, 2017

Protected equipment list May 05, 2017

Equipment Apparent Cause Evaluation for PER 959856

CR 1379519

Clearance 3-TO-2018-0001 Section 3-001-0004

OPL171.228, Electro-Hydraulic Control Logic, Revision 6

OPL171.230, Electro-Hydraulic Control (EHC) System: Hydraulics, Revision 4

OPL171.010, Main Turbine, Revision 13

ESG116.001C, Electro-Hydraulic Control (EHC) System, Revision 0

FSAR Chapter 7.11, Pressure Regulator and Turbine-Generator Control

FSAR Chapter 11.2, Turbine Generator

50.59 package for CR 1379519

ODMI for CR 1379519

Unit 3 Cycle R18 Outage Safety Plan, Revision 0

IP 71111.15

Procedures:

OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking,

Revision 24

3-SR-3.8.1.9(3C), Diesel Generator 3C Emergency Load Acceptance Test, Revision 23

0-AOI-57-1A, Loss of Offsite Power (161 and 500KV)/Station Blackout, Revision 107

Drawings:

3-47E811-1, Flow Diagram Residual Heat Removal System, Revision 72

Other Documents:

CR 1344119

PDO for CR 1344119

PDO for CR 1349343

CR 1341458 Level 1 Evaluation (RCA) Report

CR 13799519

FSAR Chapter 14.10.1, Events Resulting in a Nuclear System Pressure Increase

TS Bases 3.3.1.1

ODMI for CR 13799519

OPL171.228, Electro-Hydraulic Control Logic, Revision 6

OPL171.230, Electro-Hydraulic Control (EHC) System: Hydraulics, Revision 4

NDQ0074880118, Evaluation of LPCI Flow to Reactor Pressure Vessel (RPV) with Failed Open

Min-Flow Bypass Valve, Revision 6

MDQ0074920028, System Requirements Calculation for Residual Heat Removal (RHR)

System, Revision 6

FSAR Table 6.5-2, ECCS Equipment Capacity Assumed in LOCA Analysis

NDQ099920100006, Diesel Frequency Variation Evaluation, Revision 0

MDQ0074920113, Documentation of RHR Pump Discharge Test Flow Rates and System Test

Pressure, Revision 0

GE Letter BF 3-7413, Long Term Containment Cooling Requirements - Browns Ferry Unit 3,

dated February 27, 1976

TVA Letter, Additional Information Requested by NRC Concerning RHR Pump Protection

Against Operation in Excess of Design Runout, dated July 21, 2016

CR 1382124

BFN-VTD-G080-0771, Operation and Maintenance Instructions for General Electric NUMAC

Two-Out-of-Four Logic Module, Revision 1

CR 1382150

Browns Ferry Unit 3 Cycle 18 Core Operating Limits Report, Revision 2

ANP-3413P, Browns Ferry Unit 3 Cycle 18 Plant Parameters Document, Revision 0

Past Operability Evaluation for CR 1389131

IP 71111.18

Procedures

3-SIMI-47B, Electro-Hydraulic Control System Scaling and Setpoint Documents, Revision 41

SII-3-XX-47-204.3, Electro-Hydraulic Control System Condenser Vacuum (Turbine Exhaust)

Transmitter Calibration and Functional Test, Revision 0

Other Documents

DCN 69424, Condenser Vacuum Pressure Switches, Revision A

DCN 72342, Modify EHC Software, Revision A

PMTI-72342-03, Install Condenser Vacuum Transmitters and Provide Power Dependent Trip

Signals, Revision 2

BFN-VTD-G080-3095, General Electric Instructions - Allowable Exhaust Pressure Operation,

Revision 2

Turbine Backpressure Evaluation dated May 2016

CR 1399171

IP 71111.19

Procedures

3-SR-3.3.8.1.3 (3EB), Unit 3 4kV Shutdown Board 3EB Loss of Power Logic System Functional

Test. Revision 0008

ECI-0-000-MOV009, Testing of Motor Operated Valves, Revision 46

CCI-0-XI-00-019, Electrical Indicators, Revision 13

BFN-3-INVT-256-0001, Replace ECCS Inverter, PM Job Plan 500126964, Revs. 1, 2

ECI-0-000-BKR008, Testing and Troubleshooting of Molded case Circuit Breakers and Motor

Starter Overload Relays, Revision 0107

3-SI-4.7.A.2.g-3/3a, Primary Containment Local Leak Rate Test Reactor Feedwater Line A:

Penetration X-9A

PMTI-71673-001, Emergency High Pressure Make-up Pump Testing

WO Instructions BFN-3-BKR-211-03EC/012 Test MJ (52 Aux Switch) Switch Normal Feeder

Breaker 1338

3-SR-3.6.1.3.5(RHR II), RHR System MOV Operability Loop II, Revision 32

ECI-0-000-MOV009, Testing of Motor Operated Valves Using Viper 20, Revision 44, 47

0-TI-579(MOV), Motor Operated Valve Data Evaluations, Revision 0

3-SR-3.1.7.3, Standby Liquid Control System Enriched Sodium Pentaborate Solution

Concentration, Quantity Calculation, and ATWS Equivalency Calculation, Revision 50

3-SI-3.1.7.6, Standby Liquid Control System ATWS Equivalency Calculation for Newly

Established Pump Flow Rate, Revision 1

Drawings

3-45E766-18, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E768-3, Emergency Equipment Schematic Diagram Diesel Generator 3B

3-45E766-21, Wiring Diagram 4160V Shutdown Auxiliary Power Schematic Diagram

3-45E766-24, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E766-3, Wiring Diagram 4160V Shutdown Aux Power Schematic Diagram

3-45E724-7, Wiring Diagram 4160V Shutdown BD 3EB Single Line

3-45E68-4, Emergency Equipment Schematic Diagrams

10-11-748, Outline 5KVA inverter 250 VDC 120VAC 10 60HZ

20-113501, Schematic 5KVA Inverter 250VDC 120VAC 10 60HZ

Other Documents

WO 118373618, 118850673, 119428686, 119455067, 118369003, 119460702

IST Evaluation 18-3-IST-074-672 dated March 23, 2018

MDQ3074920442, MOV 3-FCV-74-73, Operator Requirements and Capabilities, Revision 7

IST Evaluation 18-3-IST-073-669 dated March 11, 2018 MDQ3073920407, MOV 3-FCV-73-2,

Operator Requirements and Capabilities, Revision 8

CR 1307478, 1257769, 9563525

Apparent Cause Evaluation Report for PER 956352, Revision 1

MDQ0063900083, Standby Liquid Control System Flow Analysis for ATWS Requirements,

Revision 8

MDQ0063920470, Standby Liquid Control System - Boron 10 Requirements, Revision 5

IP 71111.20

Procedures

0-OI-57A, Switchyard and 4160V AC Electrical System, Revision 163

NPG-SPP-03.21, Fatigue Rule and Work Hour Limits, Revision 20

NPG-SPP-14.1, Fitness-For-Duty and Fatigue Management, Revision 16

3-GOI-200-2A Primary Containment Entry, Revision 3

3-OI-74, RHR System Checklists for Heatup, Initiation of and Loss of Shutdown Cooling

3-GOI-100-3A, Refueling Operations (RX Vessel Disassembly and Floodup), Rev 68

Other Documents

Unit 3, Cycle R18, Outage Safety Plan

IP 71111.22

Procedures

3-SR-3.8.1.9, (3B OL) Unit 3 EDG load acceptance test,

3-SR-3.8.1.9(3B OL) - DG 3B Emergency Load Acceptance Test with Unit 3 Operating,

1-SR-3.5.1.6 (RHR II) Quarterly RHR System Rated Flow Test Loop II

3-SR-3.1.7.7 SLC System Functional Test-Pump

MCI-0-063-VLV001, Maintenance of Fired and Non-Fired SLC Explosive-Actuated Valve Units,

Revision 0025

3-SIMI-92B, Neutron Monitoring System Scaling and Setpoint Documents, Revision 0016

NESSD 3A-092-0001-00-06, Site Engineering Setpoint and Scaling Document Cover Sheet

3-SR-3.3.8.1.3 (3EB) 4KV Shutdown Board 3EB Loss of Power (LOP) Logic System Functional

Test, Revision 8

Other Documents

CR 1372616

WO 118491281

WO 118099436

WO 118491353

IP 71124.01

Procedures

NISP-RP-2, Radiation and Contamination Surveys, Revision 0

NISP-RP-4, Radiological Posting and Labeling, Revision 0

NISP-RP-5, Access Controls for High Radiation Areas, Revision 0

NPG-SPP-05.1, Radiological Controls, Revision 9

NPG-SPP-22.300, Corrective Action Program, Revision 10

RCDP-17, Radiological Postings, Revision 1

RCI-1.2, Radiation, Contamination, and Airborne Surveys, Revision 37

RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Revision 89

RCI-33, Diving Operations on the Refuel Floor, Revision 15

Data

RWP 17220021, U2 FIN Maintenance Activities

RWP 18390039, U3R18 Refuel Floor Dryer Replacement Activities

RWP 18380142, U3R18 Drywell Undervessel Maintenance Activities

RWP 18380032, U3R18 Drywell Carpenter Support Activities, Revision 1

RWP 18370011, U3R18 Reactor Building General Maintenance, Revision 0

Radiological Survey M-20180218-31, U3 Drywell 563

Radiological Survey M-20180219-18, U3 Drywell 563 Posted Ladder Access to LHRA on Top

Radiological Survey M-20180225-7, U3 RXB 565 A & C RHR Hx Update Survey

Radiological Survey M-20180220-22, U3 RXB 519 Under Torus Travel Path Survey

Radiological Survey M-20180224-22, U3 RXB 565 General Area

Radiological Survey M-20180227-56, U3 RXB 565 General Area

Radiological Survey M-20180228-30, U3 DW 550 Sub-pile Room

Radiological Survey M-20180228-15, U3 DW 550 Sub-pile Room

Radiological Survey M-20170919-20, U2 RXB 593 RWCU Pump Room

Radiological Survey M-20170922-2, U2 RXB 593 RWCU Pump Room

Radiological Survey M-20180225-14, U3 RXB 664 Refuel Floor Steam Dryer Diving Activities

Air Sample Record 18-20166-5, U3 Drywell Under Vessel GA

Air Sample Record 18-20183, U3 Drywell 550 Subpile GA

Air Sample Record M-20180301-39, 550 Subpile Room

Air Sample Record 18-20171-6, U3 Refuel Floor Divers Station

Air Sample Record 18-20107, U1 RXB 593 1A RWCU Pump Room

Six Month Inventory and/or Leak Test, August 3, 2017

U3 non-fuel inventory of Spent fuel Pool, July 14, 2017

ALARA Plan 18-0030, U3R18 Outage Steam Dryer Replacement

Tritium Activity Worksheet, U3 RCS, February 20, 2018

Alpha Level 2 and 3 Data Spreadsheet

CAP Documents

Self-Assessment BFN-RP-SSA-18-001

CR 1329077

CR 1326911

CR 1388425

CR 1390579

CR 1388749

CR 1283906

CR 1296485

CR 1291073

CR 1287739

CR 1295731

CR 1329533

IP 71124.08

Procedures, Guidance Documents, and Manuals

NPG-SPP-05.9.1, Radioactive Material/Waste Shipments, Revision 4

NPG-SPP-22.000, Performance Improvement Program, Revision6

NPG-SPP-22.300, Corrective Action Program, Revision 10

RCI-43, Radioactive Material Control, Revision 10

RWI-001, Administration of the Radioactive Material and Radwaste Packaging and

Transportation Program, Revision 12

RWI-005, Radwaste Routines, Revision 13

RWI-111, Storage of Radioactive Waste and Materials, Revision 24

RWI-156, Packaging Radioactive Material and Radioactive Waste, Revision 2

RWTP-101, 10 CFR 61 Waste Characterization, Revision 2

RWTP-102, Use of Casks, Revision 2

0-PCP-001, Process Control Program Manual (PCP), Revision 4

Records and Data

Certificate of Completion, Energy Solutions DOT/NRC Radioactive Waste Packaging,

Transportation and Disposal Training [for three qualified shippers], Various Dates

Design Change, DCN 72581, Activate LLRW Trash Module 1 to Allow Storage of Old Steam

Dryers from EPU Modifications, September 28, 2017

List of Design Changes and Temporary Modifications/ Alterations of Radwaste System since

March 1, 2016, Undated

List of Abandoned Rad Waste Processing Equipment, Undated

Radiological Survey # M-20180216-14, 10CFR37 LLRW Yard Rad Material Storage Area

Update, February 16, 2018

RCI-43, Att. 4 List of Radioactive Material Storage Areas Located Outside the Main RCA,

February 23, 2018

Radioactive Shipment Logs, Browns Ferry Nuclear Plant Radioactive Material Shipment Logs

for Calendar Years 2016, 2017, and 2018 thru January, Various

CFR Part 61 Data Waste Stream Characterization - Scaling Factors for CWPS 2016,

April 17, 2018

CFR Part 61 Data Scaling Factor Analysis-DAW 2016, April 17, 2016

CFR Part 61 Data Scaling Factor Analysis-RWCU 2016, April 17, 2016 10 CFR Part 61 Data

Scaling Factor Analysis-Thermex 2016, April 17, 2016

CFR Part 61 Data Waste Stream Characterization - Scaling Factors for CWPS 2017,

June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-DAW 2017, June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-RWCU 2017, June 8, 2017

CFR Part 61 Data Scaling Factor Analysis-Thermex 2017, June 8, 2017

Shipping Records

Shipment ID # 160107, NCMD Samples (Type A), January 3, 2016

Shipment ID # 160814, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

August 25, 2016

Shipment ID # 170201, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

February 3, 2017

Shipment ID # 171109, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

November 16, 2017

Shipment ID # 171213, Resin Shipment Type B (U) Package, LSA-III (Category 2 quantity),

December 15, 2017

Corrective Action Program (CAP) Documents

CR 1147306, 1170976, 1184022, 1219274, 1248641, 1249712, 1255883, 1307846

Self-Assessment, BFN-RP-SSA-18-001, Radiation Hazards and Transportation

December 12, 2017

IP 71151

Procedures

NPG-SPP-02.2, Performance Indicator Program, Revision 10

Desktop Guide for Identification and Reporting of NEI 99-02 Performance Indicators for

Occupational Exposure Control Effectiveness

Other Documents

Electronic Dosimeter Alarm Report, April 1, 2017 - February 12, 2018

Reactor Coolant System Leakage logs from January 1, 2017 to December 31, 2017

CR 1344430, 1330667, 1356696

IP 71152

Procedures

NEDP-8.2, Technical Evaluation for Procurement of Safety Related and Quality Related

Materials, Items, and Services, Revision 2

Other Documents

CR 1387156, 1384874, 1381298, 1375811, 1370091, 1287517, 1273615, 1262776, 1209499,

1171982, 1139533, 1133786, 1130256, 1088344, 1049856, 1038699, 1007813, 993114,

988512, 985013, 984499

Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Extended Power

Uprate Related to Amendment Nos. 299, 323, and 283 to Renewed Facility Operating License

Nos. DPR-33, DPR-52, and DPR-68 Tennessee Valley Authority Browns Ferry Nuclear Plant,

Units 1, 2, and 3 Docket Nos. 50-256, 50-260, and 50-296

WO 111044065

CR 1341458, 1357076, 1347334, 1359556

Level 1 Evaluation (RCA) Report for CR 1341458

PEG Package CYJ557C-UPGR

TVA Central Lab Services Technical Report AU27033

PEG Package 1781269-BFNX0

CNI Corrective Action Report 17-33

IP 71153

Procedures

3-AOI-100-1, Reactor Scram, Revision 65

3-AOI-100-1, Attachment 1, Scram Report dated January 10, 2018

Drawings

3-47E610-47-1, Mechanical Control Diagram Turbine, Revision 6

3-47E610-47-4, Mechanical Control Diagram Turbine, Revision 21

Other Documents

NRC Inspection Report 2017-001, Section 1R15

CR 1265552

IP 71004

Procedures

0-TI-140, Monitoring Program for Flow Accelerated Corrosion, Revision 7

IEP-200, Qualification and Certification Requirements For TVA Inspection Services

Organization (ISO) Nondestructive Examination (NDE) Personnel, Revision 16

N-UT-26, Ultrasonic Examination For Wall Thinning Conditions, Revision 30

NPG-SPP-09.7.2, Flow Accelerated Corrosion Control Program, Revision 3

Drawings

1-47E801-2, Flow Diagram Main Steam, Revision 5

1-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 19

1-47E804-1, Flow Diagram Condensate, Revision 26

1-47E805-1-1, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 5

1-47E805-1-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 2

1-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 5

2-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 0

2-47E804-1, Flow Diagram Condensate, Revision C

2-47E805-1, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

2-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

3-47E802-1, Flow Diagram Extraction Steam, Revision 0

3-47E803-1, Mechanical Flow Diagram Reactor Feedwater, Revision 0

3-47E805-2, Flow Diagram Heater Drains Vents & Miscellaneous Piping, Revision 0

Other Documents

BFN-ENG-SSA-18-001, BFN Flow Accelerated Corrosion (FAC) NRC Inspection Readiness

Self-Assessment, dated October 10, 2017

DS-M4.2.1, Flow Accelerated Corrosion Program Methods, Revision 10

EP-2015-0033-01-TR, Browns Ferry Nuclear Plant Unit 1 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-02-TR, Browns Ferry Nuclear Plant Unit 1 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 1

EP-2015-0033-03-TR, Browns Ferry Nuclear Plant Unit 2 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-04-TR, Browns Ferry Nuclear Plant Unit 2 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 0

EP-2015-0033-05-TR, Browns Ferry Nuclear Plant Unit 3 FAC System Susceptibility Evaluation

(SSE) Update, Revision 0

EP-2015-0033-06-TR, Browns Ferry Nuclear Plant Unit 3 FAC Susceptible Non-Modeled (SNM)

Analysis Update, Revision 0

Letter 15-0195-LR-001, Letter Report - Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Summary Tables for the Effect of Extended Power Uprate on Flow Accelerated Corrosion,

Revision 0

NCO 040006083, Commitment Completion Form dated October 1, 2012

EPRI State of the Fleet Assessment Tennessee Valley Authority - Browns Ferry Nuclear Plant

dated November 5, 2015

TVA Flow Accelerated Corrosion (FAC) Fleet SelfAssessment dated 6/2017

CR 974071, This SR is to document gaps found in the attached U1 FAC CHECWORKS review

CR 1081644, Evaluate Flow Accelerated Corrosion (FAC) OE from Davis-Besse (INPO Event

Report Level 3)

CR 1115734, A leak was found on the 2C3 FW Heater immediately following the December

2015 outage

CR 1219434, Through wall leak found on elbow downstream of 1-FCV-5-71 valve

CR 1380327, NRC Identified: Evaluate FAC Program SNM Technical Reports for applicability of

S3 susceptibility

Enclosure