IR 05000259/2020010

Triennial Inspection of Evaluation of Changes, Test and Experiments Baseline Inspection Report 0500259/2020010 and 05000260/2020010 and 05000296/2020010

ML20174A300
Person / Time
Site:	Browns Ferry
Issue date:	06/19/2020
From:	James Baptist NRC/RGN-II/DRS/EB1
To:	Jim Barstow Tennessee Valley Authority
References
IR 2020010
Download: ML20174A300 (11)

Text

June 19, 2020

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - TRIENNIAL INSPECTION OF EVALUATION OF CHANGES, TESTS AND EXPERIMENTS BASELINE INSPECTION REPORT 05000259/2020010 AND 05000260/2020010 AND 05000296/2020010

Dear Mr. Barstow:

On May 8, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant, Units 1, 2, and 3. On June 18, 2020, the NRC inspectors discussed the results of this inspection with Mr. Steven M. Bono and other members of your staff. The results of this inspection are documented in the enclosed report.

No findings or violations of more than minor significance were identified during this inspection.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety

Docket Nos. 05000259 and 05000260 and 05000296 License Nos. DPR-33 and DPR-52 and DPR-68

Enclosure:

Distribution via LISTSERV

ML20174A300____________

SUNSI Review

Non-Sensitive Sensitive

Publicly Available Non-Publicly Available

OFFICE RII DRS RII DRS RII DRS R II DRS

NAME G. Ottenberg M. Riley T. Su J. Baptist

DATE 06/19/2020 06/19/2020 06/19/2020 06/20/2020

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Numbers:

05000259, 05000260 and 05000296

License Numbers:

DPR-33, DPR-52 and DPR-68

Report Numbers:

05000259/2020010, 05000260/2020010 and 05000296/2020010

Licensee:

Tennessee Valley Authority (TVA)

Facility:

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

Location:

Athens, AL 35611

Inspection Dates:

May 04, 2020 to May 08, 2020

Inspectors:

G. Ottenberg, Senior Reactor Inspector

M. Riley, Reactor Inspector

T. Su, Reactor Inspector

Approved By:

James B. Baptist, Chief

Engineering Branch 1

Division of Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a triennial inspection of evaluation of changes, tests and experiments baseline inspection 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.

List of Findings and Violations

No findings or violations of more than minor significance were identified.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000260/2020010-01 Water Hammer Analysis for Unit 2 HPCI Turbine Exhaust Piping 71111.17T Open

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

71111.17T - Evaluations of Changes, Tests, and Experiments Sample Selection (IP Section 02.01)

The inspectors reviewed the following evaluations, screenings, and/or applicability determinations for 10 CFR 50.59 from May 4, 2020, to May 8, 2020.

(1) Evaluation DCN 70515, Replace Traveling Water Screens for Unit 1, 2, and 3, Rev.

C, Revision 4

(2) Evaluation DCP 72175, Unit 1-Main Steam Alternate Leakage Treatment (ALT)

Pathway Modification, Rev. 0, Revision 0

(3) Evaluation DCN 71987, Replace Unit 1 HPCI Vacuum Breaker Valves, Rev. A, Revision 0

(4) Evaluation DCN 71988, Replace Unit 2 HPCI Vacuum Breaker Valves, Rev. A, Revision 1

(5) Evaluation DCN 71865, Replace U3 HPCI Vacuum Breaker Valves and Modify Existing Piping, Rev. B, Revision 2

(6) Screen DEC 72833, MS Alternate Leakage Treatment Pathway Valve Vibration Tie-Back Support, Revision 3

(7) Screen SCN 72378, Increase Relief Valve Setpoints to Limit Undesired Actuations, Revision A

(8) Evaluation DCN 71212, Unit 2 Replace Existing Obsolete 50KV SOLA Type I&C Bus Voltage Regulators, Rev. A, Revision 0

(9) Screen DCN 71802, Create ECP to Install SLC Pump Suction Accumulator, Revision A

(10) Screen DCN 72238, Install Viscoelastic Dampers on Main Steam Lines, Revision A

(11) Screen DCN 70510, Replace Unit 2 DIV I and II HDR Inverters with Suitable Replacements, Revision A

(12) Screen DCN 70819, Isolation fuses for EDG Protective Relaying Circuits, Rev. A, Revision 0

(13) Screen DCN 72675, Torque Switch TS-17 Will be Eliminated from the Valve Closing Circuit Logic, Revision 0

(14) Screen BFN-18-216-01, BFN Unit 2 Steam Leak Detection Upgrade, Revision 0

(15) Evaluation DCN 71507 Upgrade BFN U3 Foxboro System, Rev. 3, Revision. 0

(16) Evaluation FE 1342280, Compensatory Measure for FE 1342280, Revision. 3

(17) Evaluation DCN 69532, Replace Unit 1/2 & Unit 3 Emergency Diesel Generator Governors, Rev. A, Revision. 3

(18) Screen DCN 69626, Replace Obsolete SOR Pressure Switch in Various Systems, Rev. A, Revision. 0

(19) Screen DCN 69710, Replace Obsolete MCR Recorder with Westronics Paperless Recorders, Rev. A, Revision 1

(20) Screen DCN 71265, Reconfigure Valve Control to Prevent Spurious Operation, Rev.

A, Revision 1

(21) Screen 2-AOI-47-3, Rev 22, Revised reactor power level from 30% to 26R RTP to support EPU, Revision

INSPECTION RESULTS

Unresolved Item (Open)

Water Hammer Analysis for Unit 2 HPCI Turbine Exhaust Piping URI 05000260/2020010-01 71111.17T

Description:

Engineering Change (EC) 71988 implemented a modification to the Unit 2 high pressure coolant injection (HPCI) turbine exhaust line and HPCI vacuum breaker subsystem to minimize backflow from the suppression pool to the HPCI exhaust line following a HPCI turbine trip. Updated final safety analyses (UFSAR) Section 6.4.1 states that the function of the vacuum breakers is to prevent intermittent negative pressure in the exhaust piping from pulling water out of the torus and causing a water hammer problem. The UFSAR also states that the HPCI system piping material is composed of carbon steel and is designed to USA Standard (USAS) B31.1.0-1967, Power Piping, which required in section 101.5, Dynamic Effects, that impact forces caused by all external and internal conditions (e.g., water hammer)be considered in the piping design.

The purpose of calculation MDQ0020732016000535, "Analysis of Browns Ferry Nuclear Unit 2 HPCI Turbine Exhaust Steam Line Transient Loads Following HPCI Turbine Exhaust System Plant Modifications," Revision 1, was to enable analysis of water hammer induced piping loads in the exhaust lines following the plant modifications to ensure the exhaust lines could handle the stresses upon restarting. Section 5.3 of the calculation determined HPCI could be restarted automatically and reach full speed as soon as 20 seconds following a HPCI turbine trip. This is due to a potential single failure of the HPCI turbine overspeed trip device. UFSAR Section 6.4.1 states that the HPCI system is automatically shutdown on turbine overspeed and is capable of automatically restarting if an injection signal is received and the shutdown signal no longer exists.

During the review of the EC package, the team identified that MPR Report 0048-0052-RPT-001, "Independent Technical Review on Browns Ferry Unit 2 HPCI Turbine Exhaust Piping Proposed Modification," Revision 1, stated that the calculation used a non-conservative assumption of 2% water volume in the exhaust piping as an input into the water hammer analysis and recommended that a transient analysis be performed to determine the actual amount of water in the exhaust line at the time of HPCI turbine restart. This transient analysis was documented in calculation 0048-0053-CALC-001, HPCI Turbine Exhaust Siphoning Transient, Revision 0, and determined that a water volume of up to approximately 43.8%

would exist in the exhaust piping at the time of restart. The team identified that this was a discrepancy from the 2% water volume used in calculation MDQ0020732016000535, which is used as an input into the piping stress analysis, CDQ0020732016000533, "Browns Ferry Nuclear Plant Summary of Piping Analysis N1-273-01R," Revision 4, that is used to conform to the requirements of USAS B31.1.0.

This unresolved item is being opened to determine if a violation of 10 CFR 50.59 exists. In order to make this determination, the team will need to determine whether the change would have resulted in exceeding code allowances or other applicable stress or deformation limits set in USAS B31.1.0 for the HPCI exhaust piping.

Planned Closure Actions: The licensee plans to provide an analysis detailing how the piping stresses resulting from the original 2% water volume in calculation MDQ0020732016000535 bounds the piping stresses resulting from the water volume used in Calculation 0048-0053-CALC-001, due to the differences in assumptions used in both calculations. The team will need to review the analysis to verify that the stresses caused by a HPCI restart on the turbine exhaust piping would not result in exceeding code allowances or other applicable stress limits set in USAS B31.1.0.

Licensee Actions: The licensee determined there was reasonable assurance that the HPCI system could perform its intended function until the analysis can be completed.

Corrective Action References: Condition Report

EXIT MEETINGS AND DEBRIEFS

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

• On June 18, 2020, the inspectors presented the triennial inspection of evaluation of changes, tests and experiments baseline inspection results to Mr. Steven M. Bono and other members of the licensee staff.

• On May 8, 2020, the inspectors presented the Initial Inspection Results Debrief inspection results to Steven M. Bono and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.17T Calculations

0048-0053-CALC-001

HPCI Turbine Exhaust Siphoning Transient

Rev. 0

CDQ0020732016000533 Browns Ferry Nuclear Plant Summary of Piping

Analysis N1-273-01R

Rev. 4

ED-Q-0254-880142

FUSE PROGRAM - 125Vdc Boards A, B, C, D, 3A, 3B,

3C, 3D

Rev. 7

ED-Q2000-870048

25Vdc System Short Circuit Calculations

Rev. 9

MDQ002073201600535

Analysis of Browns Ferry Nuclear Unit 2 HPCI Turbine

Exhaust Steam Line Transient Loads Following HPCI

Turbine Exhaust System Plant Modifications

Rev. 1

MDQ0068930029

MINIMUM PIPE WALL THICKNESS AND

CORROSION ALLOWANCE FOR REACTOR WATER

RECIRCULATION (RWR) SYSTEM

Rev. 5

MDQ2073910103

MOV 2-FCV-73-44 Operator Requirements and

Capabilities Calculation

Rev. 12

NDQ099920010019

Ex-Containment Removal Coefficients for Alternative

Source Term Analyses

Rev. 6

W81930916002

1/3-PS-71-1A, 1B, 1C and 1D Setpoint and Scaling

Calculation

Rev. 4

Corrective Action

Documents

CR 1341458, PER

200863, CR 1038747,

CR 1415269

Corrective Action

Documents

Resulting from

Inspection

CR 1606544

Detail in FSAR for ALT Leak Path

05/07/2020

CR 1606578

Evaluate BFN-18-216 50.59 Screening for NEI 01-01

05/07/2020

CR 1606604

During NRC 50.59 Inspection the need to review SCN

2738 screening criteria was identified.

05/07/2020

CR 1606616

Revise MDQ0020732016000535 to Clarify

Methodology

05/07/2020

CR 1606630

Revise 50.59 screen and Evaluation to incorporate

screen wash strainers

05/07/2020

CR 1606686

DCN 69710 Replace MCR Recorders. The Dry-well

effluent recorder was an analog to digital replacement

and the 50.59 was not screened in.

05/07/2020

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

CR 1606688

Revise the 10 CFR 50.59 Review for DCN 68532 to

incorporate a discussion of the failure mode for the

surge suppressor and capacitors

05/07/2020

Drawings

0-47W490-2

Mechanical Service Water, Air, Fire Protection

Rev. 8

0-47W491-7

Mechanical Service Water, Air, & Fire Protection

Rev. 14

0-47W491-9

Mechanical Service Water, Air, & Fire Protection

Rev. 3

1-47BD400-43

DC 72175, Sheets 1 and 2

Rev. 0

1-47E225-100

Harsh Environmental Data Drawing Series Index,

Notes and References

Rev. 8

1-47E225-110

Harsh Environment Data EL 565.0

Rev. 5

1-47E831-1

Flow Diagram Condenser Circulating Water

Rev. 41

1-47E850-1

Flow Diagram Fire Protection & Raw Service Water

Rev. 27

1-47E850-2

Flow Diagram Fire Protection & Raw Service Water

Rev. 31

2-45E714-1

Wiring Diagram 250 DC Reactor MOV BDS Schematic

Diagram

Rev. 16

2-45E714-4

Wiring Diagram 250V DC Reactor MOV BD 2B

Schematic Diagram

Rev. 36

2-47E610-73-1

Mechanical Control Diagram - HPCI System

Rev. 56

3-47E812-1

Flow Diagram High Pressure Coolant Injection System

Rev. 73

3-47E820-2

Flow Diagram Control Rod Drive Hydraulic System

Revs. 19

and 21

3-47E820-2-APPJ

Appendix J Testing Boundary for Control Rod Drive

Hydraulic System

Rev. 4

3-47E820-2-ISI

ASME Section XI Control Rod Drive Hydraulic System

Code Class Boundaries

Rev. 14

961111MA

Relief Valve Drawing

Rev. 0

Engineering

Changes

DCN 51607

Respan of Tech Spec Instruments for EPU

Rev. A

PIC 70629A

Revise Proposed Diesel Generator Governor Logic for

Emergency Fast Start

Rev. A

Engineering

Evaluations

BFN0EQ-IPS-003

Static-O-Ring Pressure Switches

Rev. 3

EWR18PROG064241

Review and Approval SOR Test Report 9058-102 Rev.

11/15/2018

Miscellaneous

Safety Evaluation of GE Topical Report NEDC- 31858P, Revision 2, "BWROG Report for Increasing

03/03/1999

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

MSIV Leakage Limits and Elimination of Leakage

Control Systems," September 1993

BROWN FERRY NUCLEAR PLANT, UNITS 2 AND 3 -

ISSUANCE OF AMENDMENTS REGARDING LIMITS

ON MAIN STEAM ISOLATION VALVE LEAKAGE

(TAC NOS. MA6405 AND MA6406)

03/14/2000

BROWNS FERRY NUCLEAR PLANT (BFN) -UNITS 1,

2, AND 3 -LICENSE AMENDMENT -ALTERNATIVE

SOURCE TERM

07/31/2002

Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Proposed Technical Specification Change To Revise

The Leakage Rate Through MSIVs - TS-485

11/22/2013

8001206-EMI-1

EMI/RFI Qualification Report for Engine Governor

Control System Upgrade

Rev. 0

EWR17PROJ073080

HPCI Restart Transient

Rev. 2

MPR Report 0048-0052-

RPT-001

Independent Technical Review on Browns Ferry Unit 2

HPCI Turbine Exhaust Piping Proposed Modification

Rev. 1

SR-473

2-FCV-73-44 Weak Link Analysis

Rev. 2

SS-E18.15.01

Requirements for Digital Systems(Real-Time Data

Acquisition and Control Computer Systems)

Rev. 9

TVA-BFN-TS-405

BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS

1, 2, AND 3-SUPPLEMENTAL INFORMATION

ASSOCIATED WITH RESPONSE TO

REQUEST FOR ADDITIONAL INFORMATION (RAI)

RELATED TO TECHNICAL SPECIFICATIONS (TS)

CHANGE NO. TS-405 - ALTERNATIVE SOURCE

TERM (AST) (TAC NO

S. MB5733, MB5734, MB5735).

08/24/2004

TVA-BFN-TS-436

BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 -

TECHNICAL SPECIFICATION (TS) 436 - INCREASED

MAIN STEAM ISOLATION VALVE

(MSIV) LEAKAGE RATE LIMITS AND EXEMPTION

FROM 10 CFR 50, APPENDIX J

07/09/2004

US14525-EMIR/RFIR-

0NS-0001

TVA Browns Ferry Unit 2 and 3 Digital Controls

Upgrade Electromagnetic and Radio Frequency

Rev. A

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Interference Report

Procedures

1-AOI-100-1

Reactor Scram

Rev. 24

1-OI-68

Reactor Recirculation System

Rev. 40

2-AOI-100-1

Reactor Scram

Rev. 113

2-OI-68

Reactor Recirculation System

Rev. 158

3-AOI-100-1

Reactor Scram

Rev. 71

3-OI-68

Reactor Recirculation System

Rev. 99

NPG-SPP-09.3

Plant Modifications and Engineering Change Control

Revs. 24,

2, and 33

NPG-SPP-09.3 IP-ENG-

001

Standard Design Process

Rev. 1

Work Orders

119872578, 113122016,

118008166, 118299683,

113122016, 114128873,

20136104, 118031889,

118035236, 119538467,

119873861, 118491608

PM 126676 Evaluation

Disassembly, inspection, cleaning and refurbishment of

actuator

PM126675 Evaluation

Disassembly, inspection, cleaning and refurbishment of

actuator