IR 05000259/2018010

NRC Design Bases Assurance Inspection (Team) Report 05000259/2018010, 05000260/2018010, and 05000296/2018010

ML18180A356
Person / Time
Site:	Browns Ferry
Issue date:	06/28/2018
From:	Marvin Sykes NRC/RGN-II/DRS/EB1
To:	James Shea Tennessee Valley Authority
References
IR 2018010
Download: ML18180A356 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION une 28, 2018

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC DESIGN BASES ASSURANCE INSPECTION (TEAM) REPORT 05000259/2018010, 05000260/2018010, AND 05000296/2018010

Dear Mr. Shea:

On May 18, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Plant, Units 1, 2, and 3, and the NRC inspectors discussed the results of this inspection with Mr. D. Hughes and other members of your staff. Additional inspection results were discussed during a teleconference held on June 20, 2018, with Mr. C.

Mitchell and other members of your staff. The results of this inspection are documented in the enclosed report.

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

However, inspectors documented a licensee-identified violation which was determined to be of very low safety significance in this report. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC resident inspector at the 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/

Marvin D. Sykes, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-259, 50-260, 50-296 License Nos. DPR-33, DPR-52, DPR-68

Enclosure:

Inspection Report 05000259/2018010, 05000260/2018010, and 05000296/2018010

Inspection Report

Docket Numbers: 50-259, 50-260, 50-296 License Numbers: DPR-33, DPR-52, DPR-68 Report Numbers: 05000259/2018010; 05000260/2018010 and 05000296/2018010 Enterprise Identifier: I-2018-010-0052 Licensee: Tennessee Valley Authority (TVA)

Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3 Location: Athens, AL 35611 Inspection Dates: April 30, 2018, to May 18, 2018 Inspectors: G. Ottenberg, Senior Reactor Inspector (Lead)

C. Franklin, Reactor Inspector N. Morgan, Reactor Inspector M. Riley, Reactor Inspector H. Leake, Contractor J. Zudans, Contractor Approved By: M. Sykes, Chief Engineering Branch 1 Division of Reactor Safety 1 Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring TVAs performance by conducting a Design Bases Assurance Inspection (Team) 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. Additional items are summarized in the table below. A licensee-identified non-cited violation is documented in report section 71153.

List of Findings and Violations No findings were identified.

Additional Tracking Items Type Issue Number Title Report Section Status LER 05000296/2017-002-00 4kV Shutdown Board 71153 Closed Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedure (IP) 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, performed walk downs, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

===71111.21M - Design Bases Assurance Inspection (Team)

The inspectors evaluated the following components, permanent modifications, and operating experience during the weeks of April 30 to May 4, 2018, and May 14 to May 18, 2018.

Component ===

(1) Unit 3 A and B Standby Liquid Control (SLC) Pumps [3-PMP-63-6A, 3PMP-63-6B]

a) Material condition and configuration (i.e., visual inspection during a walkdown)b) Normal, abnormal, and emergency operating procedures c) UFSAR, Design Criteria, and other design basis and licensing basis document accuracy d) Environmental qualification classification e) Protection against seismic events f) Component testing adequacy and trending results g) Maintenance effectiveness h) Consistency between station documentation (e.g., procedures, design criteria) and vendor specifications i) Translation of vendor specifications j) System isolation valve condition and monitoring (non-single failure proof)k) Recent major modifications to the system l) Condition reports associated with SLC pumps and system containment isolation valves m) Various hydraulic and net positive suction head calculations associated with this system n) Observation of a pump quarterly inservice test for pump 6A o) Latest system health report p) Review of previously identified potential non-conservative Tech. Spec for this system

(2) High Pressure Coolant Injection (HPCI) Steam Line Inboard and Outboard Valves [3-FCV-73-002, 1-FCV-73-003]

a) Material condition and configuration (i.e., visual inspection during a walkdown)b) Normal, abnormal, and emergency operating procedures c) Environmental qualification classification d) Modifications performed (permanent/temporary/minor)e) Methods and inputs for determining actuator output capability, low end of setup window, maximum allowable thrust/torque f) Review of system hydraulic calculations g) Operations monitoring of area temperatures h) Maintenance effectiveness i) Consistency between station documentation (e.g., procedures) and vendor specifications j) Translation of vendor specifications k) Testing and recent test results (Surveillance, IST, GL 96-05 Periodic Verification, Containment Isolation)l) Component health, corrective maintenance records, and corrective action history

(3) Unit 2 Anticipated Transient Without Scram (ATWS)- Recirculation Pump Trip (RPT)

Circuitry a) Material condition and configuration (i.e., visual inspection during a walk-down)b) Emergency operating procedures c) Control logic design d) Maintenance effectiveness e) Component health reports, corrective maintenance records, and corrective action history f) Surveillance testing and recent test results g) Electrical calculations

Component- Large Early Release Frequency (LERF) (1 Sample)

(1) Unit 2, A 250 Volt (V) Reactor Motor-Operated Valve (RMOV) Board a) Material condition and configuration (i.e., visual inspection during a walkdown)b) Operating procedure c) Maintenance effectiveness d) Component health reports, corrective maintenance records, and corrective action history e) Consistency between station documentation (e.g., procedures) and vendor specifications f) Bus loading and voltage calculations g) Overcurrent protection and coordination h) Protective device selection and settings i) Short circuit capacity j) Surveillance testing

Permanent Modification (4 Samples)

(1) Reflect Updated Flood Levels for Feedwater and HPCI Steam Line Breaks [DCN-72384]

(2) Respan of Tech Spec Instruments for EPU [DCNSTG- 68463-05]

(3) Replace Potential Transformers (PT) Fuses Located Inside 480V Shutdown Board 1A/1B,2A/2B, 3A/3B [DEC 72449]

(4) Replace Existing RHRSW Pumps With New Pumps that Have Closed Impellers, Upgraded Materials and a Different Bowl Assembly [DCN- 71313]

Operating Experience (2 Samples)

(1) NRC Information Notice (IN) 2018-02: Testing and Operations-Induced Degradation of 3-Stage Target Rock Valves

(2) IN 2016-09, Recent Issues Identified When Using Reverse Engineering Techniques in the Procurement of Safety Related Components OTHER BASELINE ACTIVITIES

71153 - Follow-up of Events and Notices of Enforcement Discretion Licensee Event Reports

The inspectors evaluated the following licensee event report which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:

(1) Licensee Event Report (LER) 05000296/2017-002-00, 4kV Shutdown Board Potential Transformer Primary Fuses Do Not Coordinate with Secondary Fuses

INSPECTION RESULTS

Licensee Identified Non-Cited Violation Report Section 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: The Browns Ferry Nuclear Plant, Unit 3, Renewed Facility Operating License, DPR-68, License condition 2.C(7) required, in part, that TVA Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c)... Specifically, 10 CFR 50.48(c)incorporated by reference National Fire Protection Association Standard 805 (NFPA 805),and NFPA 805 section 2.4.2.2.2, Other Required Circuits, required in part, Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

Contrary to the above, since June 22, 2016, when the NFPA 805 requirements went into effect, the licensee did not implement and maintain in effect all provisions of the approved fire protection program, because the licensee did not correctly evaluate circuits that share common power supply for their impact on their ability to achieve nuclear safety performance criteria in accordance with NFPA 805.

Significance: The team evaluated the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued May 2, 2018, and determined the finding to be of very low safety significance (Green) because the finding was related to post fire safe shutdown- circuit failure modes and affects, was not a low degradation rating, adversely affected the ability to reach and maintain hot shutdown/hot standby or safe and stable conditions using the credited safe shutdown success path, and was evaluated using the licensees risk-based evaluation, which indicated a CDF of less than 1E-6, and the evaluation result was accepted by an NRC senior reactor analyst.

Corrective Action Reference: CR

EXIT MEETINGS AND DEBRIEFS

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

• On May 18, 2018, the inspectors presented the design bases assurance (team)inspection results to Mr. D. Hughes, Site Vice President, and other members of the licensee staff.

• On June 20, 2018, additional inspection results were presented to Mr. C. Mitchell and other members of the licensees staff.

DOCUMENTS REVIEWED

Calculations

0L-084-0002-00-03, 0-L-84-2, Setpoint and Scaling Document, Rev. 3

0L-084-0013-00-05, 0-L-84-13 Setpoint and Scaling Document, Rev. 5

3L-003-0058A-00-03, 3-L-3-58A Setpoint and Scaling Document, Rev. 3

CDQ0009992014000270, Seismic Qualification Analysis for the Sulzer Supplied RHRSW

Pumps, Rev. 2

CDQ0009992015000324, Qualification of Seismic Restraints for Replacement RHRSW Pumps,

Rev. 3

CD-Q0999-890077, Detuning the BFNP RHRSW-EECW Pumps, Rev. 8

CDQ2433871714, Water Impact Analysis on Torus Shell Due to Feedwater Line Break, Rev. 1

ED-N2003-920049, Setpoint and Scaling Calculation for 2-P-3-204A, -204B, -204C, and -204D,

Rev. 7

EDQ0057920034, 4.16KV and 480V Busload, Voltage Drop and Short Circuit Calculation, Rev.

114

EDQ0057920034, 4.16KV and 480V Busload, Voltage Drop and Short Circuit Calculation, Rev.

114

EDQ0211880138, 4kV SD BDs Fuse Calculation/Evaluation, Rev. 38

EDQ024820020042, 250V DC Unit Batt Load Study, VD, SC, and Batt Capacity for

LOCA/LOOP, Station Blackout, and NFPA 805 Analysis for Unit/Shutdown, Rev. 79

EDQ0281880139, Fuse Program - 250V DC Reactor RMOV Boards 2A, 2B, & 2C, Rev. 7

EDQ0999870066, Control Circuit Voltage Drop Calc. for Typical Control Relay, Rev. 33

EDQ2000870054, CCVD Calculation - 250V DC Circuits - Unit 2, Rev. 40

EDQ2000870550, Cable and Bus Protection/Breaker/Fuse Coordination for 250V DC Systems,

Rev. 48

EDQ2574920145, Degraded Voltage Analysis, Rev. 1

EDQ2999880715, Thermal Overload Heater Calculations - Motor Operated Valves, Rev. 45

EWR No. 16PROJ02325, Clemson Engineering Hydraulics Report 597-16, dated May 2016 has

been reviewed by TVA Design Engineering and is approved for Use, Rev. 1

Enclosure

MDQ0000672013000125, Evaluation of EECW Component Flow Rates With the Most Limiting

Pump Configuration, Rev. 2

MDQ000999201200083, JOG MOV Periodic Verification Classification, Rev. 23

MDQ002320100019, RHRSW System Hydraulic Analysis for Units 1, 2 & 3 RHR Heat

Exchangers, Rev. 5

MDQ0063900083, Standby Liquid Control System Flow Analysis for Requirement, Rev. 8

MDQ0067870245, Emergency Equipment Cooling Water Available NPSH for EECW Dedicated

RHRSW Pumps, Rev. 4

MDQ0067930028, EECW System Pressure Drop - Multiflow, Rev. 8

MDQ0999980001, MOV Calculation Input Parameter, Rev. 33

MDQ107320020092, MOV 1-FCV-073-0003, Operator Requirements and Capabilities, Rev. 5

MDQ199920020033, Reactor Building High Energy Line Break Mass and Energy Release for

Environmental Analysis, Rev. 1

MDQ3073920407, MOV 3-FCV-073-0002, Operator Requirements and Capabilities, Rev. 8

NDQ0000880019, Flood Level Inside Torus Room, Corner Rooms and HPCI room Due to

Feedwater Line Break in the Main Steam Valve Vault, Rev. 6

NDQ0009992016000621, Flood Level Inside the HPCI Room Due to a HPCI Steam Line Break,

Rev. 0

NDQ0090930050, Drywell Radiation Monitor Readings Corresponding to LOCA with Various

Percentages of Clad Damage (NUMARC EALs), Rev. 11

NDQ0999880020, Flood Levels in the Main Steam Vault and Tunnel Resulting from a

Feedwater Line Break, Rev. 6

NDQ09999920076, Flooding in the Reactor Building Due to Breaks in the RWCU System, Rev.

NDQ3999910035, Summary of Harsh Environmental Conditions for Browns Ferry Unit 3, Rev.

Drawings

0-45B721-1, Fuse Tabulation, Rev. 67

0-45E702, Wiring Diagram Battery Board 2, Panel 1-7 Single Line, Sheet 1, Rev. 65

0-47W462-3, Mechanical Standby Liquid Control System, Rev. 7

1-47A370-73-30, Limit Switch Development and Calve Thrust Requirements: 1-FCV-73-03,

Rev. 4

1-47B370-2, Mechanical Motor Operated Valves Testing Requirements, Rev. 8

1-47E225-100, Harsh Environmental Data Drawing Series Index, Notes and References, Rev. 6

1-47E858-1, Flow Diagram RHR Service Water System, Rev. 72

1-47E859-1, Flow Diagram Emergency Equipment Cooling Water, Rev. 98

1-47W1392-217, Fire Protection - 10CFR50.49 Appendix R Penetration Seal Tabular Drawings

El. 565.00 Details A & B, Rev. 3

14X20BXVCM 2 Stage, Sulzer RHR Service Water Pump Curve, 12/13/2012

1-730E927, Sheet 10, Elementary Diagram, Primary CNTMT ISLN SYS, Rev. 14

2-45B720, Circuit Breaker Settings, 250V DC Reactor MOV B, Sheet 20, Rev. 3

2-45E670, ECCS Division II Analog Trip Units Schematic Diagram, Sheet 19, Rev. 21

2-45E712, Wiring Diagram, 250V Reactor MOV BD 2A, Sheet 1, Rev. 40

2-45E714, 250V DC Reactor MOV BD 2A Schematic Diagram, Sheet 2, Rev. 31

2-45E714, 250V DC Reactor MOV BD 2A Schematic Diagram, Sheet 3, Rev. 1

2-45E719, Wiring Diagram 4160V Rpt Boards Single Line, Sheet 2, Rev. 12

2-45N2641, Wiring Diagram Unit Control Board Panel 9-3, Sheet 3, Rev. 17

2-45N2641, Wiring Diagram Unit Control Board Panel 9-3, Sheet 3, Rev. 17

2-47E225-113, Harsh Environmental Data El 593.0, Rev. 3

2-730E927, Sheet 10, Elementary Diagram, Primary CNTMT ISLN SYS, Rev. 18

Enclosure

2-730E929, Elementary Diagram, Automatic Blowdown System, Sheet 1, Rev. 24

2-730E929, Elementary Diagram, Automatic Blowdown System, Sheet 1, Rev. 24

3-45E670-25, Wiring Diagram ECCS Division I Analog Trip Units Schematic Diagram SH-1,

Rev. 11

3-45E763-12, Wiring Diagram 4160V Unit Aux Power Schematic Diagram, Rev. 7

3-45E763-20, Wiring Diagram 4160V Unit Aux Power Schematic Diagram, Rev. 6

3-47A367-73-2, Motor Operated Valve 3-FCV-73-02 Thrust/Torque Requirements, Limit Switch

Development & MOV Data, Rev. 1

3-47E225-125, Harsh Environmental Data El 593.0, Rev. 8

3-47E3610-85-2, Mechanical Control Diagram CRD Hydraulic System, Rev. 16

3-47E610-23-1, Mechanical Control Diagram RHR Service Water System, Rev. 23

3-47E610-67-3, Mechanical Control Diagram Emergency Equipment Cooling Water System,

Rev. 7

3-47E854-1, Flow Diagram Standby Liquid Control System, Rev. 14

3-47E858-1, Flow Diagram RHR Service Water System, Rev. 33

3-47E859-1, Flow Diagram Emergency Equipment Cooling Water, Rev. 42

3-730E927, Sheet 10, Elementary Diagram, Primary CNTMT ISLN SYS, Rev. 10

PCS100143776-01, Pump Cross Section, 5/13/2015

PGA100143776-01, General Arr. And Section Illustration, Rev., RHRSW, 7/31/2014

Sulzer RHRSW Pump Curve 74338, 12/12/2013

Procedures

0-GOI-200-1, Freeze Protection Inspection, Rev. 87

0-GOI-300-1/ATT-9, Attachment 9 Unit 3 Reactor Building Operator Round Logs, Rev. 255

0-GOI-300-3, General Valve Operations, Rev. 142

0-OI-57D, DC Electrical System, Rev. 167

0-SI-4.5.C.1(C1), RHRSW Pump C1 IST Group A

0-SI-4.5.C.1(C1-COMP), RHRSW Pump C1 IST Comprehensive

0-SIMB-84B, Containment Atmospheric Dilution System Scaling and Setpoint Documents, Rev.

0-TI-230V, Vibration Program, Rev. 18

0-TI-345(RHRSW), RHRSW Pump Curve Data Acquisition, Rev. 6

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

10CFR50.65, Rev. 51

0-TI-360, Containment Leak Rate Programs, Rev. 43

0-TI-395, Breaker Testing and Maintenance Program, Rev. 12

0-TI-444, Augmented Inservice Testing Program, Rev. 12

0-TI-579(SLC), SLC System Baseline Data Evaluation, Rev. 2

1-ARP-9-7C, Panel 9-7, 1-XA-55-7C, Rev. 27

1-SR-3.5.1.1(HPCI), Maintenance of Filled HPCI Discharge Piping, Rev. 9

2-ARP-9-7C, Alarm Response Procedure, Revs. 10 and 25

2-SR-3.8.6.2(2), Quarterly Check of 250 Volt Main Bank Number 2 Battery, Rev. 18

3-EOI Appendix-3A, SLC Injection, Rev. 3

3-OI-63, Standby Liquid Control System, Rev. 24

3-SI-4.4.A.1, Standby Liquid Control System Functional Test, Rev. 60

3-SR-3.1.7.7, Standby Liquid Control System Functional Test, Rev. 30

3-SR-3.3.3.1.4(G), Verification of Remote Position Indicators for High Pressure Coolant

Injection System Valves, Rev. 7

BTI-EEB-TI-28, Setpoint Calculations, Rev. 11

DS-M18.2.21, Motor Operated Valve Thrust and Torque Calculations, Rev. 24

Enclosure

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

Starter Overload Relays, Rev. 107

ECI-0-000-MOV009, Testing of Motor Operated Valves, Rev. 48

ECI-0-248-BAT001, Equalize Charging the 250 Volt Main Bank Batteries, Rev. 12

EPI-0-000-BKR003, General Electric Type AK-15/25 Circuit Breakers and Switch Gear

Maintenance, Rev. 84

G-50, Torque and Limit Switch Settings for Motor-Operated Valves, Rev. 9

IP-ENG-001, Standard Design Process, Rev. 0

MCI-0-063-ACC001, SLC Accumulator Maintenance, Rev. 18

NEDP-2, Design Calculation Process Control, Rev. 22

NEDP-2.1, Electrical Auxiliary Power System Modeling, Rev. 0

NEDP-2.2, Electrical Power System Analysis Calculations, Rev. 1

NEDP-8.6, Reverse Engineering Services for Mechanical Items, Rev. 1

NPG-SPP-01.1, Administration of Standard Programs & Processes, Standard Department

Procedures, and Business Practices, Rev. 9

NPG-SPP-01.2, Administration of Site Technical Procedures, Rev. 17

NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitor, Trending, and Reporting -

CFR 50.65, Rev. 3

NPG-SPP-06.2, Preventive Maintenance, Rev. 12

NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 2

NPG-SPP-09.0.1, Conduct of Systems Engineering and Equipment Reliability, Rev. 5

NPG-SPP-09.1, ASME Code and Augmented Programs, Rev.9

NPG-SPP-09.16.1, System, Component and Program Health, Rev. 11

NPG-SPP-09.18.13, Equipment Failure Trending, Rev. 0

NPG-SPP-09.2, Equipment Environmental Qualification Program, Rev. 10

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Revs. 9 and 28

Pump Test, Rev. 5

Quarterly Pump Test, Rev. 10

Design Basis Documents

BFN-50-7023, Residual Heat Removal Service Water System, Rev. 28

BFN-50-7063, Standby Liquid Control Design Criteria Document, Rev. 20

BFN-50-7067, Emergency Equipment Cooling Water System, Rev. 24

BFN-50-7068, Reactor Water Recirculation System, Rev. 14

BFN-50-7073, High Pressure Coolant Injection System, Rev, 31

BFN-50-7200C, 250 V DC Power Distribution System, Rev. 8

BFN-50-7200D, 480V AC Auxiliary Power System, Rev. 12

BFN-50-728, Physical Independence of Electrical Systems, Rev. 25

Plant Modifications

2449, Replace potential transformers (PT) fuses located inside 480V Shutdown Board

1A/1B,2A/2B,3A/3B, Rev. A

2704, Replace potential transformer (PT) fuses in 4kV boards, Rev. 1

DCN 51198, U1 Recovery Reactor Building Mechanical Lead System 073, Rev. A

DCN 68463A, Respan of Tech Spec Instruments for EPU, February 8, 2018

DCN 71802, Create ECP to Install SLC Pump Suction Accumulator, Rev. A

DCN71313, Replace Existing RHRSW Pumps With New Pumps that Have Closed Impellers,

Upgraded Materials and a Different Bowl Assembly, Rev A

DEC 72400, SDP - Standby Liquid Control (SLC) Pump Discharge Accumulator Gauge/Valve

Assembly, Rev. A

EDC 51074, Revise documents to reflect AST requirements see page 2 (attached), Rev. A

Enclosure

EQV 71281, Modify DC MOVs to Improve Valve Stroke Time Margin, Rev. A

SCN 72371, Replace Stem/Stem Nut and Increase Packing Load to Reduce the Potential for

Coasting into the Backseat During the Opening Stroke, Rev. A

Miscellaneous Documents

298689-BFNC0, Accumulator, Typical, QA 1, Hydraulic, Stainless Steel, For HCU, General

Electric, CBD755P, Rev. 0

- 900 Lb. Double Disc Gate Valve, Rev. 7

150617-CKJ710EG0, Reverse Engineering Documentation Package - Flowserve/BWIP Valve

Wedges CWR233F, Rev. 1

2-EOI-1A, ATWS RPV Control Flow Control, May 16, 2018

336655-BFNG, Vendor Documents (ASME Sec. XI, & Drawings) Review and Approval,

CRM066A, Clyde Union Pump Co., EWR 11PGM063214 and EWR11MEB063215

BFN0EQ-CSC-001, Conax Corporation - Electric Conductor Seal Assembly, Rev. 22

BFN0EQ-HS-001, Cutler-Hammer - Handswitches, Rev. 16

BFN0EQ-TBK-001, Terminal Blocks, Rev. 25

BFN-50-727, Environmental Qualification, Rev. 6

BFN-VTD-A391-0340, Instruction Manual for Anchor Darling

BFN-VTD-B260-0140, Sulzer Installation Operation and Maintenance Manual

BFN-VTD-F990-0165, Maintenance Manual for Anchor/Darling Double-Disc Gate Valves 2 - 1/2

And Over, Rev. 4

BFN-VTD-G080-1440, Instructions Installation and Maintenance of 7700 Line Motor Control

Center, Rev. 2

BFN-VTD-L200-0010, Vendor Technical Manual for Limitorque Valve Operators, Rev. 21

BFN-VTD-L200-0220, Limitorque Type SMB 10 CFR Part 21 Notifications Maintenance

Updates and Technical Updates, Rev. 17

BFN-VTD-L200-0260, Limitorque-SMB Series/SB Installation and Maintenance Manual, Rev. 7

BFN-VTD-U055-0020, Union Pump Co., Instruction Manual for Installation and Operating TD-60

Power Pump Class, Rev. 4

BFN-VTD-V085-0020, Instruction Manual for Operation and Maintenance of Velan Manual

Bolted Bonnet Gate, Globe Stop Check and Check Valves, Rev. 4

BFN-VTD-V085-0120, Forged Steel Valves Operation and Installation Manual, Rev. (Valves 1/4 -

2)

BFN-VTM-G080-1550, Vendor Technical Manual for General Electric 250-V Reactor MOV

Boards, Rev. 13

Browns Ferry Nuclear Plant Motor Operated Valve Trending Report (U3C17), Dated 7/25/2016

Browns Ferry Residual Heat Removal Service Water Pump Specification, Rev. 2

DS-M18.14.1, Design Standard for Environmental Qualification of Electrical Equipment in Harsh

Environments, Rev. 4

GE-NE-A22-00125-42-01, Task T0609: Standby Liquid Control System, Rev. 1

Health Report, System 63, Standby Liquid Control, Oct. 16 to March 17

NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitoring, Trending and

POM1001437760010-01, Rev. 0

Reactor Recirculation System 068A/B Second Period System Health Report, 2017

Relay-Control-Electromechanical, Preventative Maintenance Database, April 24, 2018

Reporting - 10CFR50.65, Rev. 3

Reporting - 10CFR50.65, Rev. 51

RHRSW Pumps, Vibration and Flow trends and graphical representations, New Pumps, Various

Dates.

SAR Change Package 17-159, October 5, 1998

SAR Change Package 72642-01, November 27, 2017

Enclosure

SCR1147D, 1 PH AC - UPS System 2B, Rev. B

Standby Liquid Control System, trends and graphical representations, Various dates.

Switchgear Low Voltage Preventative Maintenance Database, April 23, 2018

TP16-1-112, Recommendations to Resolve Flowserve 10CFR Part 21 Notification Affection

Anchor Darling Double Disc Gate Valve Wedge Pin Failures, Rev. 4

VTD-G080-0450, Testing and Maintenance of General Electric Molded Case Circuit Breakers,

Rev. 2

VTD-G080-1250, General Electric Molded Case Circuit Breakers Application and Selection,

Rev. 4

VTD-G080-1300, E 100 Line Molded Case Circuit Breakers Enclosure Compensated, Rev. 1

VTD-G080-1310, General Electric Molded Case Circuit Breakers F225 Line Ambient

Compensated Front Adjustable Trip, Rev. 3

VTD-G080-2825, Application and Selection of General Electric Molded Case Circuit Breakers,

Rev. 0

Condition Reports

20418 0946406 1180357 956352 1192007

22478 1139620 1193742 1143066 1023943

1141448 1142007 1142474 1161842 1183598

1145657 1193771 1237903 CAQR CAQR

BFP880466 BFP871034

114949416 1315529 1375227 1269428 1251754

1171957 1167863 1101819 1062524 1251754

1199002 1251656 1354129 1354138 1395412

283642-005 1334072 1379185 0854255

Work Orders

09-718856-000 116781658 117483118 95-12472-00 02-003308-000

1147366 0117656136 118795542 119435030 00-003899-000

113740237 116781673 117651692 119415324 119343263

116781647 117479103 114693925-T 06-713118-000 117845209

117557873 118020640 118065081 118308521

117845209 95-012472 09-718856 11469392

118435926 118435929 09-725975-000 115752639

Condition Reports Written Due to this Inspection

CR 1410526, [NRC Identified] - Unsecured Cart U3 RB El. 565' near East Bank HCUs

CR 1410844, NRC identified concern with tarps on RHRSW roof grating

CR 1410846, NRC identified concern with housekeeping in the D RHRSW Pump Room

CR 1410988, 2A RMOV BD DBAI walkdown action

CR 1411570, Documentation issue discovered during the 2018 DBA

I.

CR 1413186, Calculation EDQ0211880138 R36 - DBAI Inspection

CR 1413537, DBAI Inspection - Bend Radius Issue for Battery Room 2

CR 1414957, EQ devices in containment not previously evaluated for chemical effects

CR 1414959, Evaluate raising SLC injection radiation monitor criteria such that a LOCA does

not result in SLC

CR 1415132, DBAI Walkdown additional information

CR 1415217, (DBAI): Battery Charger Short Circuit contribution not included Calculation

EDQ024820020042

Enclosure

CR 1415861, (DBAI): Vendor (General Electric) Molded Case Circuit Breaker Interrupting

Ratings

CR 1415887, 2018 DBAI URI: 250V DC RMOV Board

CR 1419666, MAXIMO doesnt match calculation for EQ device

Enclosure