ML17333A626

Summary of 961022 Meeting W/Util to Discuss Former Bolts Insp & Repair Plans for 1997 Refueling Outage.List of Attendees & Viewgraphs Encl

ML17333A626
Person / Time
Site:	Cook
Issue date:	10/29/1996
From:	John Hickman NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9611040262
Download: ML17333A626 (46)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 29, 1996 LICENSEE:

American Electric Power Service Corporation (AEP)

(Indiana Michigan Power)

FACILITY:

D.C.'Cook, Unit 1

SUBJECT:

FORMER BOLTS INSPECTION AND REPAIR PLANS FOR THE 1997 REFUELING OUTAGE A meeting was held at NRC Headquarters on October 22, 1996, with the AEP and Westinghouse representatives, to discuss the licensee's plans for the inspection and replacement of the D.C.

Cook Unit 1 core barrel former bolts, during the plant's upcoming refueling outage.

During the 1994 Unit 1 refueling outage, a bolt and associated locking bar were found on the reactor vessel internal's lower core plate.

The bolt was identified as the type used to secure the core baffle former plates to the core barrel.

During a subsequent refueling outage in 1995, the reactor core was defueled and the core barrel was removed from the vessel.

An inspection of all the core barrel former bolts was performed.

A bolt missing from the top former plate was identified.

Further testing disclosed that two former bolts located on either side of the missing bolt were loose.

On September 26, 1995, the licensee met with NRR to review the root cause and safety evaluation conclusions based on the inspection results and analyses performed with support from Westinghouse (internals supplier).

Based on the safety evaluation at that time, the licensee concluded that the plant could safely operate during the upcoming cycle without replacing the three former bolts or making any other repairs.

The licensee had committed to (1) reinspect the former bolts at the next refueling outage, (2) develop and prepare repair procedures, methods and equipment, (3) monitor and analyze neutron noise and loose parts data throughout the operating cycle, and (4) keep the NRC informed of any new developments.

During this meeting, in the discussion and responses to NRR questions, the licensee and Westinghouse representatives provided an update on the root cause evaluation and analyses and presented a description of the procedures, techniques, equipment and redesign of replacement former bolts that had been mc KK CSPV 9611040262 96i029 PDR ADOCK 050003i5 8

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The licensee outlined its plans for a reinspection of the core barrel former bolts and implementation of the repair and replacement of the three bolts, during the next refueling outage.

The next refueling outage is scheduled for February 1997.

Docket Nos: 50-315 John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Enclosures:

1. List of attendees

2. Slides cc w/encls:

See next page

0 developed.

The licensee outlined its plans for a reinspection of the core barrel former bolts and implementation of the repair and replacement of the three bolts, during the next refueli'ng outage.

The next refueling outage is scheduled for February 1997.

r Docket Nos: 50-315 Original signed by:

John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Enclosures:

1. List of attendees

2. Slides cc w/encls:

See next page DISTRIBUTION w~encls Docket File PUBLIC PD3-3 R/F ACRS OGC e-mail FHiraglia RZimmerman JRoe EAdensam EJordan (JKR)

GMarcus BMcCabe JCaldwell, RIII FGrubelich WBurton DOCUMENT NAME:

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Indiana Michigan Power Company CC:

Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Hichigan 48913 Township Supervisor Lake Township Hall P.O.

Box 818 Bridgman, Michigan 49106 Al Blind, Site Vice President Donald C.

Cook Nuclear Plant 1 Cook Place Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire

Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.

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Washington, DC 20037 Mayor, City of Bridgman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 State Capitol Lansing, Michigan 48909 Drinking Water and Radiological Protection Division Hichigan Department of Environmental guality 3423 N. Hartin Luther King Jr Blvd P. 0.

Box 30630 CPH Hailroom

Lansing, Michigan 48909-8130 Donald C.

Cook Nuclear Plant Mr. Steve J.

Brewer Indiana Hichigan Power Nuclear Generation Group 500 Circle Drive

Buchanan, Hichigan 49107 Hr. E.E. Fitzpatrick, Vice President Indiana Hichigan Power Nuclear Generation Group 500 Circle Drive

Buchanan, HI 49107

MEETING ATTENDEES FOR OCTOBER 22, 1996 D.C.

COOK UNIT 1

FORMER BOLTS INSPECTION AND REPAIR PLANS MEETING NAME John B. Hickman Frank Grubelich Steve Brewer Frank R. Pisarsky John C. Matarazzo David A. Howell Melita P.

Osborne AFFILIATION NRC, Project Directorate 111-3 NRC, Mechanical Engineering Branch AEP, Regulatory Affairs AEP, Production Engineering Westinghouse, Engineering Westinghouse, NSD Westinghouse, NSD ENCLOSURE 1

~ UNIT 1 BARREL-FORMER BOLT DISCUSSION

~ INTRODUCTION AND BACKGROUND STEVE BREWER, AEP

~ ROOT CAUSE EVALUATIONUPDATE MELITAOSBORNE, WESTINGHOUSE

~ BOLT REPAIR PROCEDURE DAVE HOWELL, WESTINGHOUSE

~ 1997 U1 REFUELING OUTAGE PLANS FRANK PISARSKY, AEP

UPPER CORE BARREL

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A6 A5 A4 THERMAL SHIELD PLAN VIEW, BARREL FORMER BAFFLE BOLTING

o UNIT 1 1994 REFUELING OUTAGE

>> LOOSE BOLT FOUND ON LOWER CORE PLATE

>> BOLT IDENTIFIEDAS BARREL FORMER BOLT

>> WESTINGHOUSE PERFORMED SAFETY EVALUATION

>> DECISION TO PERFORM VISUAL INSPECTION NEXT REFUELING OUTAGE

~ UNIT 1 1995 REFUELING OUTAGE

>> FOUND MISSING BOLT LOCATION (A5)

>> FOUND TWO LOOSE ADJACENT BOLTS (A4 AND A6)

PLANT OPERATION ACCEPTABLE FOR NEXT FUEL CYCLE

~ CONSIDERED:

>> LOOSE PARTS

>> EFFECTS OF NEXT CYCLE OF OPERATION ON REMAININGBARREL FORMER BOLTS

>> FUEL ASSEMBLY GRID CRUSH

~ CONSIDERED:

>> BYPASS FLOW CHANGES

>> FLOW-INDUCED VIBRATION

>> CONTROL ROD INSERTION CAPABILITY

~ INSPECTION OF BARREL FORMER.

BOLTS NEXT OUTAGE COMMUNICATESCOPE AND TECHNOLOGY TO NRC

~ CONTINUE TO REFINE ROOT CAUSE ANALYSIS

~ DEVELOPMENT OF APPROPRIATE REPAIR APPROACH, TECHNIQUES

ROOT CAUSE ASSESSMENT FROM 1995 JA(445%4%&

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A5 h YARN'6'Y4 (Yh YhcwhYP wAvvlvPk&M4ws4 )%4/ANvwwwcNM Causes Eliminated

~ Stress corrosion:

NOT seen in examined bolts

~ Manufacturing/Assembly:

NOT supported by Westinghouse records or examination of bolts Postulated Root Cause:

~ Acceptable loads were exceeded locally thermal shield support blocks

~ A4-A6 locations

~ Mechanisms result from thermal transients and flow induced vibration

ROOT CAUSE ASSESSMENT 1996 (Cont)

Flow Induced Vibration (FIV) Analyses

~ Purpose is to determine local variations in loads for as-built configuration

~ FIV loads caused by flow of water through the internals (e.g., horizontal flowfrom nozzle changes directions as hit barrel)

~ Investigate as-built thermal shield support block configuration

~ 360'-D model of barrel, baffle and thermal shield, A-H formers

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ROOT CAUSE ASSESSMENT 1996 (cont)

Flow induced Vibration (FIV) Analyses Preliminary Results,

~ Bolts located under thermal shield support blocks have highest loads

~ Bolts located under 4 of the 8 locations have higher loads magnitude for these four locations essentially the same A4-6 are under one of these four thermal shield support blocks

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ROOT CAUSE ASSESSMENT 1996 (cont)

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~ As-built configuration {dowel pins) has negligible effect on loads

~ Highest overall loads on A former level

~ Predicted frequencies of vibration correlate with measured neutron noise data

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ROOT CAUSE ASSESSMENT 1996 (cont)

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>A Steady State Thermal Analysis

~ Purpose is to determine relative thermal growth between baffle plates, former plates, and core barrel

~ Determine steady state thermal loads on former bolts

~ Thermal loads cause bending loads in bolts

~ May cause loss of preload, allow bolt to vibrate

ROOT CAUSE ASSESSMENT 1996 (cont)

~ Same model as FIV analysis; as-built configurations

~ Plant specific thermal loads {heat generation rates, T,, T,,)

~ Analysis still in progress

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D. C. COOK BARREL TO FORMER BOLTS COMPARISON ORIGINAL BOLT REPLACEMENT BOLT Thread Effective Length of Engagement Material Spec Material Yield Strength Ultimate Strength 3/4 - 10 1.18 in.

W PDS 70041KB 347 SS 30-55 KSI 75 KSI 1"-8 0.805 in.

SA 479 316 SS Strain Hardened 65 KSI 85 KSI

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D. C. COOK BARREL TO FORMER BOLTS COMPARISON OLD BOLT NEW BOLT DIMENSIONS DIMENSIONS Length Overall:

Bolt Head Height:

Shank Length:

Shank Diameter:

Bolt Head Diameter:

3.500 3.920 0.750 0.962 2.750 2.775 0.620 0.620 1.125 1.440

D. C. COOK BARREL TO FORMER BOLTS COMPARISON NSQ?XtlWNCv'

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D. C. COOK BARREL TO FORMER BOLT LOCKING DEVICES Device Original Lock Bar New Crimp Torque to Undo 48 F.t-Lb 150 Ft-Lb

BARREL-FORMER NEW BOLT DESIGN TESTING

Purpose:

Confirm the performance ofthe newly-designed bolt with respect to the original bolt to support new bolt design Load-Deflection Measure strain in each bolt design at different loading levels when applied in a direction perpendicular to bolt longitudinal axis Comparative Fatigue Tests:

Determine bolt preload for each bolt design; monitor preload degradation throughout test Fatigue Testing willbe completed at the Westinghouse Science and Technology Center (STC)

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REPAIR EQUIPMENT

~ Elevator delivery device is used for positioning

~ End effector with EDM probe used to machine a 3.0" through hole in the thermal shield

~ EDM is qualified to remove the old bolt

~ Vacuum system used to clean up debris

~ EDM used to machine larger hole for bolt and notches for locking device

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REPAIR EQUIPMENT

~ Tapping end effector used to machine new 1"-8 threads

~ Torquing tool installs new bolt

~ Crimping tool locks bolt in place

~ Hole in thermal shield acceptable without repair

BARREL - FORMER BOLT INSTALLATIONDEMONSTRATION a Mockup willbe a simulated section ot the DC Cook reactor vessel lower internals

~ Mockup and tools willbe submerged underwater

~ The demonstration willshow the following operations

~ Positioning of the tool on the thermal shield

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BARREL - FORMER BOLT INSTALLATIONDEMONSTRATION

~ EDM of the new 1" bolt hole"

~ Tapping of the 1" bolt hole

~ Installation and torque of bolt

~ Crimping of the bolt

~ Removal of the tool from the thermal shield

• EDM of the 3" access hole through the thermal shield was previsouly demonstrated

Frank Pisarsky Engineering Manager of Mechanical Components D.C. Cook Nuclear Plant

ore arre ns ections

~ Section XI visual inspection of core support structures

~ Visual inspection of all 640 former bolt locations

~ Mechanical agitation of A Level former bolts, B Level in proximity of discrepant bolts, 56 random bolts, and H Level bolts P3, 4, 5, 6, 7, 8 8

~ Detailed visual inspections of key areas for movement/stress

~ UT measurements of gap between barrel wall and A Level former plate e Fiberscope inspection of discrepant bolt holes

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e u e ns ections

~ Visual inspection and mechanical agitation of all A (Top) Level former bolts

~ Visual inspection and mechanical agitation of B Level bolts (B1 through BB) in proximity of missing bolt holes (A4, A5 8 A6)

~ Minimum of 86 bolts inspected

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BCCWCIlt e Install 1" Bolts in holes A4, A5, A6

~ Return Core Barrel to original configuration (A4, A5, A6 3i4" to 1" bolts)

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