Text

Summary of Meeting with Browns Ferry to Discuss One Preliminary Greater than Green Finding Associated with One Apparent Violation That Was Documented in NRC Inspection Report 05000259-10-005, 05000260-10-005, and 05000296-10-005
	ML111010106
Person / Time
Site:	Browns Ferry
Issue date:	04/08/2011
From:	Eugene Guthrie; Reactor Projects Region 2 Branch 6
To:	Krich R; Tennessee Valley Authority
References
	EA-11-018 IR-10-005
	Download: ML111010106 (52)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION ril 8, 2011

SUBJECT:

PUBLIC MEETING SUMMARY- REGULATORY CONFERENCE FOR BROWNS FERRY NUCLEAR PLANT, DOCKET NO. 50-259

Dear Mr. Krich:

This refers to the meeting conducted on April 04, 2011, in Atlanta, GA. The purpose of this meeting, was to allow representatives of Tennessee Valley Authority (TVA), the licensee for Browns Ferry Nuclear Station, Unit 1, to meet with U. S. Nuclear Regulatory Commission (NRC)

personnel to discuss one preliminary Greater Than Green finding associated with one Apparent Violation that was documented in NRC Inspection Report 05000259/2010005, 05000260/2010005, and 05000296/2010005 (ML110610641). The findings dealt with the failure to establish adequate design control and perform adequate maintenance on the Unit 1 low pressure coolant injection (LPCI) outboard injection valve, 1-FCV-74-66, resulting in the valve being left in a significantly degraded condition and residual heat removal (RHR) loop II unable to fulfill its safety function.

A list of attendees and a copy of your presentation handout are enclosed.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Should you have any questions concerning this meeting, please contact me at (404) 997-4662.

Sincerely,

/Craig Kontz RA for/

Eugene F. Guthrie, Chief Projects Branch 6 Division of Reactor Projects Docket No.: 50-259 License No.: DRP-33 Enclosure: As stated cc w/encl: (See page 2)

__ML111010106 __ G SUNSI REVIEW COMPLETE OFFICE RII:DRP RII:DRP SIGNATURE JDH /RA for/ EFG /RA/

NAME CKontz EGuthrie DATE 04/08/2011 04/08/2011 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

TVA 2 cc w/encl:

K. J. Polson Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution C.J. Gannon General Manager Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution J. E. Emens Manager, Licensing Browns Ferry Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution T. C. Matthews Manager, Corporate Nuclear Licensing - BFN Tennessee Valley Authority Electronic Mail Distribution State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 30317 Montgomery, AL 36130-3017 E. J. Vigluicci Assistant General Counsel Tennessee Valley Authority Electronic Mail Distribution Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611 James L. McNees, CHP Director Office of Radiation Control Alabama Dept. of Public Health P. O. Box 303017 Montgomery, AL 36130-3017

Letter to R. from Eugene Guthrie dated April 8, 2011 SUBJECT: PUBLIC MEETING SUMMARY- REGULATORY CONFERENCE FOR BROWNS FERRY NUCLEAR PLANT, DOCKET NO. 50-259 Distribution w/encl:

C. Evans, RII L. Douglas, RII OE Mail RIDSNRRDIRS PUBLIC RidsNrrPMBrownsFerry Resource

Enclosure 1

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 1 Regulatory Conference Low Pressure Coolant Injection Valve 1-FCV-74-66 Atlanta, Georgia April 4, 2011 Enclosure 2

Agenda

Introduction Preston Swafford

Background Rob Whalen

Root Cause/Engineering Analyses Rob Whalen

Significance Determination James Emens

Performance Deficiency James Emens

Corrective Actions Rob Whalen

Long-Term Fire Strategies Rob Whalen

Closing Remarks Preston Swafford

Enclosure 2

Introduction

On October 23, 2010, Residual Heat Removal (RHR) Loop II Injection Valve, 1-FCV-74-66, did not pass flow while operations was aligning shutdown cooling at the beginning of a refueling outage

Disassembly of the valve revealed the disc separated from the stem and lodged in the seat

The disc separation from the stem resulted from an original manufacturing defect, undersized threads in the disc skirt/disc assembly

- Preliminary root cause was thought to be lack of skirt key caused disc separation, which was basis for apparent violation

- Final root cause shows that the cause of the disc separation is not a licensee performance deficiency

Based on results of extensive forensic examination, analysis, and laboratory mockup tests, we have shown conclusively that the disc would have released within an acceptable time with an RHR pump running (due to friction reduction from pressure pulsations), allowing the valve to provide functional flow

TVA is taking steps to significantly reduce risk due to fire at the Browns Ferry Nuclear Plant

- Reducing instances of Self-Induced Station Blackout (SISBO) actions

- Accelerating plant changes identified as part of NFPA 805 transition

- Changing the Safe Shutdown Instructions (SSIs) to allow the use of alternate shutdown paths

Enclosure 2

Background -

Assessment Approach

Root cause team was assembled including site and corporate expertise

Comprehensive forensics were performed to determine root cause

- Southwest Research Laboratory (weld examinations)

- Westinghouse Laboratory (valve component forensics)

- Structural Integrity (thread strength analysis, sensitivity study)

- Independent Burns & Roe metallurgist (aggregate review of forensics reports)

Performance Improvement International (PII) performed detailed analysis and laboratory testing to determine the valves capability to function in its as-found state of separation

Enclosure 2

Background -

Timeline Timeline 1968 Walworth valve purchased as an assembly from General Electric for construction of Browns Ferry Nuclear Plant (BFN),

Unit 1 December 1974 Separation/loosening of discs due to flow-induced vibration 1983 Installed modified disc with V notch trim (skirt reused)

June 2006 Replaced stem prior to BFN, Unit 1, restart due to observed stem nut damage 2007 to October 2010 Satisfactory quarterly valve stroke times based on limit switch indication, not torque March 2009 Initiated shutdown cooling passing 7,000 gpm flow through valve October 2010 Initiated shutdown cooling with no observed flow (terminated pump operation after 110 seconds)

Enclosure 2

Background -

RHR Low Pressure Coolant Injection (LPCI) System Overview 74-66 valve

Enclosure 2

Background -

RHR Shutdown Cooling Overview 74-66 valve

Enclosure 2

Root Cause -

RHR Valve Cross Section Stem Skirt OUTLET Tack Weld Disc Threads INLET V Notch Trim Flow

Enclosure 2

Root Cause -

Disc Separation Forensics (Disc/Skirt Threads)

Enclosure 2

Root Cause -

Disc Separation Forensics (continued)

Forensic examination found axial damage on the threaded connection between disc and skirt Optical Comparator Image of Damaged Threads

Enclosure 2

Root Cause Results

The manufacturer supplied an undersized skirt to disc connection male thread diameter under a 10 CFR 50, Appendix B program (Part 21 report submitted via revision to Licensee Event Report on April 1, 2011)

The valve was purchased as an assembly that would not be taken apart to perform receipt inspections

- Undersized skirt thread diameter caused the threaded connection between disc and skirt to be 38 percent of design strength

- Pressure on skirt/stem side of disc due to downstream check valve leakage and surveillance testing configuration

- Tack welds designed to prevent rotational, not axial separation

- Stem and skirt pulled away from disc in open direction

Disc was initially separated from stem/skirt before November 2008, based on Motor-Operated Valve Actuator Testing (MOVAT) data review and forensic examination

- This indicates that the valve passed normal shutdown cooling flow in a separated condition in March 2009

Enclosure 2

Root Cause Summary

NRC Inspection Report 2010-05 noted . . . the licensees failure to establish adequate design control and perform adequate maintenance on the Unit 1 outboard LPCI injection valve, 1-FCV-74-66, which resulted in the valve being left in a significantly degraded condition and RHR loop II unable to fulfill its safety function, was a performance deficiency.

Root cause analysis determined that no licensee performance deficiency existed

- No reasonable basis existed to examine threads and identify the undersized thread condition

Corrective actions discussed later in this presentation

No other root or contributing cause was identified

We will show that the valve, while being degraded, would have performed its fire safe shutdown safety function

Enclosure 2

Functionality Analysis

Industry research shows that pump-induced vibrations dramatically reduce frictional forces

Idaho National Laboratory research shows static coefficient of friction behavior for stellite valve seating surfaces 74-66 valve disc contacting surface immediately after removal (11/2/10)

Enclosure 2

Analysis of Force Balance and Coefficient of Friction Significance

A two-dimensional static analysis was performed to determine the normal force and coefficient of friction

The calculated coefficient of friction was well aligned with the method discussed in Idaho National Laboratory stellite aging research 1

An energy balance approach was used to determine the energy applied by the disc to the valve body and associated deflection from each stem stroke Idaho National Engineering and Environmental Laboratory Document, INEEL/EXT-02-01021,Results of NRC-Sponsored

Stellite 6 Aging and Friction Testing, October 2002. 14 Enclosure 2