Text

Core Support Barrel Insp,Repair & Analysis & Thermal Shield Removal
	ML20085H551
Person / Time
Site:	Saint Lucie
Issue date:	08/16/1983
From:	; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	;
Shared Package
ML20085H549	List: ML20085H551;
References
	CEN-253(F)-NP, NUDOCS 8309090649
	Download: ML20085H551 (34)
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- CEN-253(F)-NP 1

ST. LUCIE UNIT 1 CORE SUPPORT BARREL INSPECTION, REPAIR AND ANALYSIS

, AND 4' THERMAL SHIELD REMOVAL 9

August 16, 1983 Combustion Engineering, Inc.

Nuclear Power Systems Windsor, Connecticut 0309090649 B30830 PDR ADOCK 05000335 , l P PDR

AGENDA FOR PROPRIETARY MEETING WITH NRC:

ST. LUCIE 1 CORE SUPPORT BARREL INSPECTION, REPAIR AND ANALYSIS, AND THERMAL SHIELD REMOVAL I. REACTOR VESSEL INTERNALS INSPECTION RESULTS, REPAIR AND ANALYSIS '-

A. PROJECT . STATUS OVERVIEW B. CORE SUPPORT BARREL REPAIR

1. STATUS

2. INTERNALS DESCRIPTION

3. - CORE SUPPORT BARREL INSPECTION--RESUL-TS---VISUAL-

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4. NDE' INSPECTION TECHNIQUES

5. DESCRIPTION OF REPAIR OPTIONS

6. FUNCTIONAL CRITERIA FOR REPAIR C. STATUS OF FAILURE MECHANISM ANALYSIS II. THERMAL SHIELD REMOVAL A. STATUS .. . _ _

B. REMOVAL TECHNIQUES l

PROJECT STATUS OVERVIEW - TASKS COMPLETED INITIAL ASSESSMENT OF REACTOR 4/12/83 VESSEL INTERNALS REACTOR VESSEL INTERACTIONS (PTS) L-83-263 4/27/83 '

EFFECTS ON FUEL PERFORMANCE L-83-265 4/27/83 TECHNICAL SPECIFICATION P/T LIMITS L-83-280 5/3/83 LPMS AND EXCORE DETECTOR DATA L-83-345 6/7/83 REVIEW RV GAMMA HEATING SUBMITTAL L-83-367 6/23/83 EXXON RELOAD SUBMITTAL L-83-369 6/23/83 FINAL EXXON RELOAD SUBMITTAL L-83-429 7/27/83 6

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. RETURN TO SERVICE 0F THE CORE SUPPORT BARREL g STATUS

- g *. INTERNAL'S DESCRIPTION g IN5PECTION RESULTS g NON DESTRUCTIVE EXAMINATION TECHNIQUES 9 REPAIR TECHNIQUES g .

ANALYSIS ,

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STATUS OF THE RETURN TO SERVICE ACT:VITIES OF THE REACTOR INTERNALS (EXCLUSIVE OF THE CORE SUPPORT BARREL)

G , VISUAL EXAMINATION OF REACTOR INTERNALS BY UNDERWATER TELEVISION HAS BEEN COMPLETED e 'THERE IS NO EVID.ENCE OF DAMAGE TO THE REACTOR INTERNALS AND/OR THE INTERFACES WITH THE REACTOR VESSEL & FUEL ASSEMBLIES G EVALUATION OF THE REACTOR INTERNALS STRESSES CONSIDERING PLANT OPERATION WITHOUT A THERMAL SHIELD HAS BEEN COMPLETED e THE REACTOR INTERNALS S'RESSES ARE WITHIN THE ASME CODE SECTION III, SUBSECTION NG ALL0HABLE STRESSES FOR:

NORMAL'0PERATION & UPSET FAULTED CONDITION 9 e em S

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STATUS OF THE RETURN TO SERVICE ACTIVITIES OF THE CORE SUPPORT BARREL s*

(D VISUAL EXAMINATION OF CORE SUPPORT BARREL (D PRELIMINARY ASSESSMENT OF DAMAGED CORE SUPPORT BARREL STRUCTURAL INTEGRITY GD NON DESTRUCTIVE EXAMINATION OF DAMAGED AREA

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(p UNDERWATER MACHINING EQUIPMENT & MOCKUPS dp REPAIR MACHINING GD VERIFICATION OF REPAIR PACHINING 4

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VISUAL EXAMINAT:0N RESULTS 35 MM SLIDES OF NINE THERMAL SHIELD LUG LOCATIONS LUG LOCATION NUMBER OF SLIDES 7 1 9

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NON DESTRUCTIVE EXAMINATION TECHNIOUES 4 INSPECTION e VISUAL EXAMINATION (VT) BY UND,ERWATER TELEVISION o G.

EDDY CURRENT (ET) -- TO DETERMINE EXTENT OF SURFACE CRACKS e ULTRA-SONIC (UT) - 'TO DETERMINE EXTENT OF CRACKS S VERIFICATION OF REPAIR MACHINING e .

EDDY C U R R E N_T ,

9 TECHNIQUES AND ACCEPTANCE CRITERIA e ASME BOILER & PRESSURE VESSEL CODE SECTION XI, RULES FOR INSERVICE INSPECTION OF NUCLEAR POWER PLANT COMPONENTS 4

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CORE SUPPORT BARREL REPAIR MACHINING OBJECTIVES i

G t1INIMIZE REMOVAL OF CORE SUPPORT BARREL MATERIAL

$ MACHINE DAMAGED AREA 0F CORE SUPPORT BARREL TO REDUCE STRESS CONCENTRATION O MACHINING NECESSARY.TO INSTALL THE REPAIR TO LIMIT BYPASS COOLANT FLOW e 4 e em S

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CORE SUPPORT BARREL STRESS CRITERIA:

ASME CODE SECTION III, CONSIDERING THE FOLLOWING CONDITIONS:

NORMAL OPERATION AND UPSET SSE LOCA 4

9 _ CORE SUPPORT BARREL PRELIMINARY STRESS RESULTS:

e THE CORE SUPPORT BARREL STRESSES CONSIDERING OPERATION WITHOUT A THERMAL SHIELD ARE WITHIN THE ASME CODE SECTION III, SUBSECTION NG ALLOWABLE STRESSES e FATIGUE EVALUATION BASED ON MAINTAINIMG A 11 FATIGUE USAGE FACTOR <1 FOR 10 CYCLES 9

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FA f LURE MECHANI SM_ANALYS I S PPOC-RAM S DHASE 1: BASIC INFORMATION I PHASE 2: EVALUATION AND INTEGRATION OF BASIC INFORMATION O PHASE 3: IDENTIFICATION OF FAILURE PECHANISM l

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FAILURE MECHANISM ANALYSIS PROGRAM BASIC INFORMATION STATUS e HYDRAULIC LOADS CALCULATED PERIODIC AND RANDOM LOADS, e STRUCTURAL RESPONSE CALCULATED FREE VIBRATION RESPONSE CALCULATIONS ON FORCED RESPONSE IN PROGRESS.

e DESIGN, FABRICATION, FURTHER EFFORT.0N FIELD INSTALLATION.

INSTALLATION DATA e DAMAGE & VISUAL EVALUATED PHOTOGRAPHIC EVIDENCE OF INSPECTION DAMAGE WITH THERMAL SHIELD ON. WILL PERFORM SINILAR EVALUATION NOW SHIELD IS REMOVED.

e LPM & IVM DATA EVALUATED THREE SETS OF LPM DATA AND IVM DATA FROM CYCLES 1 TO 5.

e METALLURGICAL COMPLETED ELECTR0tl MICROSCOPY WORK ON EXAMINATION PORTIONS OF THE SHIELD. CONTINUING WORK ON EXAMINATION OF OTHER SAMPLES AS WELL AS SUPPORT PINS.

FAILURE MECHANISM ANALYSI'S PROGRAM DISCUSSION WILL CONCENTRATE ON PROGRESS IN:

e METALLURGICAL EXAMINATION OF PORTIONS OF THE SHIELD.

e STRUCTURAL RESPONSE CALCULATIONS -

FREE VIBRATIONS, e REDUCTION OF ADDITIONAL IVM DATA,

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METALLURGICAL EXAMINATION EXAMINATION OF PORTIONS OF THE THERMAL SHIELD, SUPPORT PINS AND LOCKING BARS. .

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EXAMINATION OF POSITIONING PINS l

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STRUCTURAL RESPONSE G FINITE ELEMENT MODELS TO COMPUTE VARIATION OF SYSTEM (COUPLED CORE SUPPORT BARREL AND THERMAL SHIELD) FREQUENCY WITH DEGRADED SUPPORT CONDITIONS.

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9 CONTINUING WORK:

CHARACTERIZATION OF TYPES OF IMPULSES. QUANTIFY AMPLITUDE AND RATES OF IMPACT.

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TVM DATA

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CYCLE DATE 2 BOC JUNE, 1978 EOC- -MA-RCHT 1979 3 BOC JUNE, 1979 EOC FE3., 1980 4 BOC MAY, 198n EOC SEPT., 1981 5 BOC DEC., 1981 APRIL _, 1982 , , , ,_

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THERMAL SHIELD REMOVAL r

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LAST THERMAL SHIELD SEGMENTS REMOVED JULY 26, 1983 l

l 0 REMOVAL TECHNIQUE THERMAL SHIELD CUT-OFF USING PLASMA ARC TORCH 9

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ML20085H551

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