ML19309B538
| ML19309B538 | |
| Person / Time | |
|---|---|
| Issue date: | 01/23/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19309B528 | List: |
| References | |
| ACRS-1658, NUDOCS 8004040256 | |
| Download: ML19309B538 (30) | |
Text
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MEETING DATE: July 10,1979 i
DATE ISSUED : January 2'3, 1980
//v/s MINUTES Or TsE ACRS SueCOMMITTEE O
MEETING ON METAL COMPONENTS JULY 10,1979 WASHINGTON, D.C.
I The ACRS Subcommittee on Metal Components met in Room 1046, 1717 H Street, N.W., Washington, DC to review the NRC research program sponsored by the General and Water Reactor Safety Research Offices.
No written corraents or requests for time to make oral statements from members of the public.have been received. Selected handouts received by the participants are attached. The office copy of the minutes contains all the handouts.
l Executive Session (0 pen)
Dr. P. Shewmon opened the meeting by stating 1.he scope of the meeting is to review the research and technical assistance programs that are funded by the NRC in preparation for a report to Congress.
C. Serpan, Chief Metallurgy & Materials Research Branch, NRC
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Mr. Serpan reviewed the activities in the primary system integrity pmgram.
These areas are listed belowl The program is conducted in four major areas.
1.
Vessel and Piping Fracture Mechanics, 2.
Irradiation Effects and Dosemetry.
Steam Generator Tube Integrity and Stress Corrosion.
3.
4.
Non-Destruct 1ve Examination,
800404oE SG
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. Metal Components The purpose of the vessel and piping fracture mechanics program is to
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and piping to assure validate the fracture mechanics analysis for vessels safe design and operation.
In this area various coments were made concernino Some of the ma.ior the environmental effects of piping and vessel degradation.
comments are listed below:
.The effects of high oxygen content in BWR coolant on startup and pH excursions is not known at this time.
.The chemistry problem thus far has not yet been scoped.
.The cyclic growth rate as predicted by the ASME code is not conservative, for both water and air environment. The difference is by a factor of about 1.5 to 2 and may be as high as 10.
In the area of piping fracture mechanics three items mentioned by Mr. Serpan were as follows:
1.
Reevaluate the cold leg break criteria.
2.
Reevaluate the criteria for pipe breaks and pipe whip.
3.
Establish piping analysis methodology and propose an updated licensing criteria.
These programs are directed toward determining piping system reliability Major coments in this area include the following:
. The reliability analysis of the piping system will account for time variaticn of stresses, such as stress changes as the, crack pmpagates, stress changes as the internal piping pressure decreases and stress changes due to varying seismic and dynamic loadings.
. The pipe break location criteria will be validated.
.The concern of piping system overdesign will be considered in the program.
.Too much effort may be expended. to generate sophisticated comouter code where sin.plistic codes now exist to determine upper bound of pipe deflections due to pipe whip.
o.
Metal Components. Data from foreign countries will be factored in the program.
. Prediction of stress corrosion crack that might exist in the structure is difficult if not impossible. The Staff also stated that to predict how fast stress corrosion crack will grow and exactly the conditions under which it will grow is difficult.
.The prediction is difficult because the tools needed to analyze this condition is not yet at hand.
If it was in the brittle fracture mechanics area, more accurate prediction could be made. Crack behavior in the elastic-plastic range is difficult, to predict.
.The approach to the piping reliability program is to assume a flaw either environmentally or mechanically induced and determining the severity o f the fl aw.
.The Subcommittee concern is to detemine how the crack is initiated and what the environmental conditions are for the crack to gmw.
.The Staff stated that rules for the design of a plant that will prevent stress corrosion cracking currently exist. There are rules for the prevention of sensitization in stainless steel and rules on limiting The Staff also stated that if the rules were the stresses.
foltowed most of these cracks, if not all of them, probably would not The Staff stated that the form of regulation covernino occur.
oxyger. content in water is not yet available.
The second area covered by Mr. Serpan is the program on irradiation effects and dosimetry. The purpose of this program is to establish valid irradiation effects trends and methods to predict fluence and embrittlement in reactors.
From the metallurgical aspect the program will establish the ductile shelf toughness of weld metali, and reevaluate the 10 CFR 50 rules on the Charpy-V, 50 ft-lb toughness criteria. The program will also validata the nrassure vessel surteillance dosimetry procedures and evaluate the validity of embrittlement saturation.
The pmgram on Charpy-V, 50 ft-lb toughness criteria will determine the ductile shelf maximum toughness of thick irradiated weld specimens by the use of Test will be perfomed at elevated temperature and the J-R curve techniques.
analysis will be made using inelastic fracture mechanics methods.
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.4 Metal Components dosimetry and embrittlement pro-The dosimentry program will update Expected from this pmgram is sel surveill.
.e.
jections for pressure ves lyze and interpret reactor a series of standards that will detemine how to ana The elements of the program are listed below:
surveillance results.
d dosimeters.
. Type of surveillance capsule specimen an ter to the pressure ves i
. Extrapolation of the dose from the dos me Displacement investigated.
.The concept of flux or fluence will beper atom or O Majo' comments in this section are listed below:
s in the fluence calculations
.This program was initiated because errorErrors in the calculatio are not acceptable.
d with the low shelf weld material ssure vessel life i
>mb ne
.The error in the calculationproperties becomes important lowest.
when toughness of the pressure vessel are ht forward by Tne embrittlement saturation pmblem was first broug 9
ittlement Their data suggest that at fluence below 10, embr Westinghouse.
d any increase fluence on the might very well saturate in power reactors an in embrittlement.
vessel would not cause a corresponding inc pressure EPRI is pursuing the program.
for testina purposes.
from various reactors having a wide range of fluences t al fluence in the The Staff stated that over the past ten years the ac u lculation by h
surveillance capsule has been much higher than a factor of 2 to 3 times.
te this type of error in i
The Staff program is attempting to reduce or elim na the future.
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. Metal Components The third area of the program deals with steam generater tube integrity and stress corrosion in order to develop an understanding of tube integrity The Staff stated and the mechanism of degradation of steam generator tubes.
that the tube integrity problem is essentially a program to satisfy the licensing Results of this program thus far indicates that a tube wastage of about board.
The Staff 40 and 60 percent of the thickness reques the tubes to be plugged.
is trying to ship a degraded steam generator (Surry 2) from its present site in Virginia to Richland, Washington, in order to perform non-destructive examination research, decontamination, etc. The Committee from the safety view point, questioned the usefulness'of this 4.5 million dollar program over the next half a dozen years.
l Mr. Serpan next discussed BWR pipe cracking specifically the Duane Arnold problems, which was caused by improper design. A program to study stress corrosion cracking in both stainless, teel and ferritic steel will be initiated.
Variation of oxygen-water chemistry and other parameters will also be studied.
Most of this work is being done by vendors and Mr. Serpan stated that the Staff need to develop an independent capability in this area in order to be able to assess ongoing programs.
The fourth area covered concerns non-destructive examination (NDE).
The
- ogram will establish improved methods for ultrasonic flaw detection.
Evaluation of continuous acoustic emission mondtoring the detection of flaws l
l and crack growth, will also be' studied.
A i
Mr. Serpan handout of his presentation is attached in Appendix A.
l detailed budget of the Metallurgy and Materials Research Branch is also included in this Appendix.
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. Meta! Components Mr. Bosnak of the Mechanical Engineering Branch next explained the inter-He stated that there is an interply action between Research and the user group.
He stated that follow-betweer his gmup and Research through the review group.
up with Research is limited because of lack of manpower.
Technical assistance programs (TAP) from DSS and DOR were discussed next.
TAP in the area of metallurgy and materials are essentially confined to the engineering branches of DSS and 00R. The following are major highlights of this discussion:
.A new material engineering pmgram has been proposed to address environmental effects of ferritic steels, including reactor vessels.
This program will also identify rational requirements to be placed on the control of secondary water chemistry.
This pmgram will study protective
.The protective coating program is new.
coating used throughout the plant and develop criteria necessary for acceptance of these various coatings in nuclear plants.
. Tearing stability in LWR piping is being performed by Dr. P. Parris of The objective of this program is to perform Washington University.
tearing stability analyses for various flaw conditions, sizes, shapes, and orientation in LWR piping and structures.
.The containment buckling behavior program will develop acceptance criteria for the design of the containment structure against buckling.
This problem arose during the FNP when the base plateform under postulated loads deflected causing the containment to be subjected to loads not accounted for in the analysis. Similar problems also exist in ice condenser plants and containments with asymetric loads.
EpRI presentation Mr. Gary Dau, Pmgram Manager for the nondestructive evaluation pmgram, summarized the activities that are under way at EPRI. Three areas of work were mentioned. The are:
- 1) improvement of code inspection techniques, 2) new methods and equipment development and 3) material pmperty measurement in order to identify materials degradation before failure occurrs. A total of 31 pro-jects have been completed or are currently underway and since the initiation of the project in 1974, a total of $50 million has been spent.
=
. Metal Components In the Basic work is being done in the NDE and acoustic emission area.
material property measurement area work is being concentrated on the measuring of residual stresses by the use of acoustic emission and x-ray defractometer techniques.
EPRI is developing a dedicated utility industry center with nondestructive The focus of this center is to provide technology transfer testing capability.
between the research laborctory to the field, which includes training, equipment ano procedures.
Mr. Danko also of EPRI covered the pmgrams in the areas of water chemistry and corrosion with specific attention on the BWR pipe cracking problem. The thrust of the program in Mr. Danko's area is to address the reliability and availability of LWR piping systems.
Mr. T. Mager, Manager of Metallurgy at Westinghouse, presented their research activities in the area of primary boundary integrity. Their programs include NDE, stress analysis, material properties, fracture mechanics, and irradiation effects on pressure vessel materials.
Some new programs for the 1979/1980 period is listed below:
. Development of a crack arrest toughness data bank for irradiated RVP mate rial s.
. Steady state radiation embrittlement of reactor vessels.
Mr. Moore, is with the nuclear power generation gmup at Babcock & Wilcox.
His presentation covered Babcock & Wilcox research programs including some results of their reactor vessel owners group pmgram.
B&W steam generators tubes have experience high cycle fatigue failures.
Mr. Moore discussed B&W efforts in performing fatigue test to determine the 1
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Metal Components failure mechanism of once-through steam generator tubes,. B&W is also involved Mr.
in developing predictive curves for irradiation induced pmperty chanoes.
Moore also discussed the weld metal chemistry problem of the reactor vessel beltline welds.
Mr. Cowan, of General Electric Co., reviewed their pmgram related to plant materials development. The top priority program is the evaluation of stress corrosion cracking of BWR materials.
Essentially, no new pmgrams have been initiated and escentiellv tie aracentation was a review of existing programs and presented some results.
.ielected handouts of his presentation are shown in Appendix B.
Mr. Ayres is a cupervisor of Special Analysis, Plant Engineering at Combustion Engineering Power Systems. He discussed two programs related on reactor to materials and mechanics. They are: 1) irradiation effects vessel saterials and 2) development of pipe break criteria.
CE maintains a developmental effort in the area of irradiation effects in reactor vessel materials to provide maximum assurance of the integrity of the RPV. Specific efforts in this area include reactor surveillance program, irradiation effects, prediction and analysis, fabrication methods, material j
monitoring and participation in the development of industry standards.
i Mr. Ayres also discussed the research work on the validity of the pipe In this pmgram, computer analyses are being performed on large break criteria.
diameter pipe of the primary pressure boundary. The purpose of this pmgram is to determine the likelihood of a safe shutdown earthquake, causing a pipe The break. The analysis includes internal pressures and seismic loadings.
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Metal Comp:nents 9-analysis assumes a circumferential through-wall-crack halfway around the pipe and located at the outside of a pipe bend radius, where the stresses due to seismic loadings are the maximum.
The result of the analysis indicates that it is not realistic to assume a complete pipe break.
In fact the analysis indicates that under the assumed pressure and seismic loadin's the assumed pipe crack remains stable and a guillotine type failure does not occur. Further work is con-tinuing in this area.
The meeting was adjourned at 7:08 p.m.
A, complete transcript of the open sessions of this meeting is NOTE:
on file at the NRC Public Document Room at 1717 "H"
- Street, N.W., Washington, DC or can be obtained from ACE Federal Reporters, Inc., 444 North Capitol Street, Washington, DC 20001 (202) 347-3700.
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ACRS SUBCOMMITTEE MEETING ON METAL COMPONENTS, EXTREME EXTERNAL PHENOMENA, AND COMBINATION OF DYNAMIC LOADS JULY 10-11,1979 WASHINGTON, DC ACRS NRC STAFF P. Shewmon, Chaiman P. Albrecht D. Okrent B. Turovlin M. Bender E. Woolridge H. Etherington A. Taboada C. Mark C. Grimes C. Siess D. Jeng Z. Zudans, Consultant R. Bosnak J. Moteff, Consultant W. Berry, Consultant M. Wechsler, Consultant ATOMIC INDUSTRIAL FORUM A. Pense, Consultant T. Pickel, Consultant J. O'Neill W. Lipinski, Consultant R. Scavuzzo, Consultant H. Corten, Consultant TENNESSEE VALLEY AUTHORITY E. Rodabaugh, Consultant E. Igne, Designated Federal Employee W. Hall R. Savio, ACRS Staff GENERAL ELECTRIC CO.
LAWRENCE LIVERMORE LAB.
N. Shirley J. Weaver R. Cowan COMBUSTION ENGINEERING CORP.
BCS T. Natan A. Williams D. Ayres B. Feamster IEAL SWEC J. Quinn C. Grochmal NUTECH TELEDYNE ENGINEERING SERVICES T. Martin J. Larson UPI WESTINGHOUSE ELECTRIC CORP.
]
E. Roby T. Mager STONE & WEBSTER BABCOCK & WILCOX A. Eckert B. Miyeh E. Siskin S. Jacobs M. Tan K. Reinschmidt M. Holley K. Moore D. King R. Klause J. Hall W. Kennedy
. ELECTRIC POWER RESEARCH INST.
J. Danko G. Dau R. Leyse KEPC0 K. Ota PUBLIC E. O'Brien.
S. Holzsweig T. Dillon B. Shitlock R. Cloud
TENTATIVE SCHEDULE COMBINED ACRS SUBCOMMITTEE MEETING METAL COMPONENTS EXTREME EXTERNAL PHENOMENA COMBINATION OF DYNAMIC LOAD JULY 10 & 11,1979 WASHINGTON D.C.
I JULY 10,1979 A__PPROXIMATE TIME MEETING OF THE JOINT ACRS SUBCOMMITTEES ON METAL COMPONENTS AND EXTREME EXTERNAL PHENOMENA
)
8:30 am I.
EXECUTIVE SESSION A.
Chairman's Opening Statement - Shewmon B. ' Members & Consultants Comments II. NRC PRESENTATION A.
NRC Research Activities
- emphasis is to be on need for research program and its priority relative to other program needs 8:40 - 11:30 1.
Metallurgy & Materials - Serpan
- include Surry Steam generator inspection program at Hanford 10:40 - 10:50
- BREAK ****************************
B.
NRC Technical' Assistance Program 11:30 - 12:15 1.
DSS - Knight 12:15 - 1:10 2.
00R - Noonan
- LUNCH *****************************
1:10 - 2:10 III. EPRI PROGRAMS 2:10 - 2:55 A.
Material & Metallurgy - Jones 2:55 - 3:40 B.
NDE - Dau
- BREAK *****************************
3:40 - 3:50
. JULY 10, 1979 (Cont'd)
APPROXIMATE TIME s.
IV. VENDOR PROGRAMS 3:50 - 4:35 A.
1-Mager 4:35 - 5:20 B.
B&W - Moore 5:20 - 6:00 C.
GE - Cowan 6:00 - 6:40 D.
CE - Ayres 6:40 - 7:15 V.
DISCUSSION OF SUBCOMMITTEE RECOMMENDATIONS 7:15 VI. ADJOURNMENT JULY ll,1979 8:30 am I.
EXECUTIVE SESSION NRC RESEARCH (Extreme External Phenomens Subcommittee)
II.
8:45 - 9:00 A.
Introductory Statement - Shao 9:00 - 10:30 B.
Mechanical Engineering - Richardson 10:30 - 11:00 C.
Structural Engineering - Bagchi D.
Research directed toward the prediction of recurrence 11:00 - 11:30 intervals for earthquakes at reactor site - Harbour 11:30 - 12:30 III. DISCUSSION OF THE JUBCOMMITTEE RECOMMENDATIONS
- LUNCH ***************************
12:30 - 1:30 MEETING OF THE ACRS SUBCOMMITTEE ON COMBINATION OF DYNAMIC LOADS 1:30 Z.
EXECUTIVE SESSION (0 pen)
A.
Chairman's Opening Statement 8.
Members and Consultants Connents 1:40 - 2:15
!!. STATEMENT BY WILLIAM KENNEDY. VP ENGINEERING. SAM III. UPDATE OF STATUS OF PLANT SHUTDOWNS AND OTHER AFFECTED PLANTS 2:15 - 2:45 2:45 - 3:15 IV. LESSONS LEARNED FROM PLANT SHUTDOWNS AND REANALYSIS
- BREAK ***********************
3:15 - 3:25 l
1
i 3-JULY 11,1979 (cont'd)
APPROXIMATE TIME 3:25 - 4:00 OPERATIONAL MONITORING OF PLANTS WITH REGARDS TO
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V.
STRUCTURAL CAPABILITY OF PIPING SYSTEMS AND ITS RESTRAINTS 4:00 - 5:00 VI. LOAD COMBIRATION Basis, philosophy, background, and future directions.
A.
B.
Current criteria 5:00 - 5:30 VII. RESEARCH/ TECHNICAL AS!! STANCE PROGRAM A.
Load Combination Failure P'ropagation -- verification of leak before B.
break criteria 5:30 - 6:00 DISCUSSION AND STATUS OF FEEDWATER LINE CRACKS VIII.
- Do these cracks suggest a lack of conservatism in the design procedures or design data used in the PWR reactor pressure vessel?
- What was the cause of the cracks and/or what are we doing to find out.
6:00 - 6:30 DISCUSSION OF SCBCOMMITTEE RECOMMENDATIONS IX.
6:30 X.
ADJOURNMENT I
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l METALLURGY & MATERIALS RESEARCH BRANCH PROGRAM AREAS o
VESSEL AND PIPING FRACTURE MECHANICS o
IRRADIATION EFFECTS AND DOSIMETRY o.
STEAM GENERATOR TUBE INTEGRITY AND STRESS CORROSION o
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VESSEL AND PIPIPG FRACTURE MECHANICS ESTABLISH VALIDATED FUNDAMENTALS OF FRACTURE AND F PURPOSE:
' PIPING TO ASSURE SAFE DESIGN AND OPERATION.
DEVELOP AND VAO DATE E/P AND TEARING INSTABILITY CRITERIA OBJECTIVES:
o VALIDATE THERMAL SH0CK AND STEAM LINE BREAK EVALUATION ME o
ESTABLISH INTEGRITY OF VESSEL HAVING LOW SHELF MATERIAL o
UPDATE ASME CODE CURVES FOR CRACK GROWTH RATE o
ESTABLISH CRACK ARREST TOUGHNESS ANALYSIS o
REEVALUATE COLD LEG BREAK CRITERIA o
REEVALUATE CRITERIA FOR PIPE BREAKS AND PIPE WHIP o
i ESTABLISH PIPING ANALYSIS METHOD AND PROPOSE UPDATE o
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IRRADIATION EFFECTS AND D0SIMETRY PURPOSE:
ESTABLISH VALID IRRADIATION EFFECTS TRENDS AND METil0DS TO PREDICT FLU AND EMBRITTLEMENT IN REACTORS.
ESTABLISH DUCTILE SHELF TOUGHNESS OF IRRADIATED WELD METALS OBJECTIVES:
o REEVALUATE 10CFR50 RULES ON CHARPY-V 50 FT-LB TOUGHNESS CRITERION o
SUPPORT FRACTURE MECilANICS TECHNOLOGIES IN CRACK ARREST AND CR o
ESTABLISH CRITERIA FOR CYCLIC IRRADIATION - ANNEALING 0F PRESSURE VE o
VALIDATE PRESSURE VESSEL SURVEILLANCE - DOSIMETRY PROCEDURES o
EVALUATE VAllDITY OF EMBRITTLEMENT SATURATION o
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3
NON-DESTRUCTIVE EVALUATION PURPOSE:
ESTABLISH IMPROVED METHODS AND CRITERIA FOR ULTRASONIC FLAW DETEC EVALUATION, AND FOR CONTINUOUS ACOUSTIC NONITORING FOR FLAW INITIATION AND
' GROWTH.
VALIDATE IMPROVED SAFT-UT FLAW CHARACTERIZATION DURING ISI.
OBJECTIVES:
o DEVELOP IMPROVED "REAL-TIME" FLAW DETECTION TECHNIQUES.
o DEVELOP IMPROVED INTERACTIVE FLAW DISPLAY SYSTEM.
o ESTABLISH PROBABILITY OF FLAW DETECTION FOR SPECIFIC SIZE FLAWS.
o REVISE AND UPDATE ISI REQUIREMENTS FOR MOST CRITICAL FLAW TYP o
DEVELOP AND VALIDATE IMPROVED EDDY CURRENT ISI FOR SG TUBES.
o DEVELOP AND VALIDATE AC0USTIC EMISSION AND INTERNAL FRICTION FOR o
CONTINUOUS MONITORING OF REACTOR STRUCTURAL INTEGRITY.
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DETAllFD BUDGET - ETAllllRGY f. P!ATERIAIS RESEARCH BRANCH (BASIC BUDGET - N0 SUPPLEE NTS)
FIN #
CONTR.
TITLE FY 79 FY 80 FY 81 VESSFl AND PIPING FRACTURE ECHANICS B6290 NSRDC J-R CURVE TESTING 122 100 SAE A4046 BCL CRACK ARREST 305 325 DOWN
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B5609 BCL COLD LEG BREAK 225 0
COMPL.
A0133 SAI PIPING RELIABILITY 0
200 UP UNDES.
DEGRADED PIPE STAB.
0 0
START A0136 SANDIA 2-PHASE JET LOADS (200) 0 330 SAE A1216 U CAL.
PIPE WHIP (150) 0 150 SAE B0119 ORNL HSST PROGRAM 2207 1890 SAE B0123 ORNL DESIGN CRIT, PIPING 145 0
COMPL.
A9026 UMD PHOT 0 ELASTIC C/A 141 0
COMPL.
B6288 WASH. U.
TEARING INSTABILITY 108 175 UP B0412 ORNL RES. ENGR IN FRG 70 80 SAE UNDES.
HYDROGEN EXPLOSIONS 0
0 START UNDES.
HYDROGEN EMBRITTLEE NT 0
0 START UNDES.
INTERACTIVE PIPE CODE O
0 START I
ORNL PRESSURIZED T/S n
a START l
3323 3250
'NOT APPROVED BY BRG (
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DETAllFD BUDGET - METALLURGY & MATERIALS RESEARCH BRANCH GASIC BUDGET - NO SUPPLEMENTS)
FIN #
- CONTR, TillI FY 79 FY 80 FY 81 IRRADIATID'N EFFECTS AND DOSIMETRY BSS28 NRL IRRAD. EFFECT IN RPV 725 800 UP B5988 HEDL SURVEILLANCE DOSIMETRY 540 500 UP B6224 NBS DOSIMETRY DATA BASE 72 90 SAME i
790 750 DOWN 2127 2140 STEAM GFNERATOR AND STRESS CORROSION CRACKING 4
B3012 GE ST. ST. SUSCEPT. TEST 235 0
COMPL.
B2097 PNL SG TUBE INTEGRITY 656 985 UP A3208 BNL SG TUBE CORROSION 210 225 SAME UNDES.
0 START UNDES.
0 0
START UNDES.
RESERVE o
a llP 1101 1210 i
'NOT APPROVED BY BRG (
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DETAllFD BUDGET - ETAIIURGY & MATERIALS RESEARCH BRANCH (BASIC BUDGET - N0 SUPPLEMENTS)
FIN #
CONTR.
TITLE EY_Z3 FY 80 FY 81 N0NDESTRbCTIVEEXAMINATION l
A4052 GARD A/E WELD MONITOR 80 0
COMPL.
l B2088 PNL A/E VESSEL MONITOR 350 400 UP B5605 SwRI SAFT-UT FOR ISI 0
81 UP l
A4047 U. MICH.
SAFT-UT DEVELOP.
237 200 UP l
B0417 ORNL EDDY CURRENT PROBES 93 114 UP B6212 DAl INTERNAL FRICTION (200) 0 200 SAME B2289 PNL INTEGRATE NDE & F.M.
274 645 SAME j
I B6327 UNDES.
IMPROVED DETECTION 0
0 START UNDES.
NEW TECHNIQUES 0
0 START 360 DOWN UNDES.
UNDES.
REQUAL. AFTER ACCIDENT 0
0 START 1034 2000 TOTAL 7600 8600
- NOT APPROVED BY BRG (
)*
i
M hS\\R. Q GENERAL ELECTRIC COMPANY NUCLEAR TECHNOLOGY DEPARTMENT REVIEW 0F Pl. ANT MATERIALS RELATED DEVELOPMENT PROGRAMS 1
7/10/79 R.L.COWAN hom. A s e
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9 4
PLANT MATERIALS DEVELOPMENT PROGRAMS _
e EVALUATION OF STRESS CORROSION CRACKING OF BWR MATERIALS,
- FUNDAMENTAL UNDERSTANDING
- PARAMETRIC STUDIES
- SHORT TERM REMEDIES
- ALTERNATE MATERIALS
- IMPROVEMENTS APPLICABLE TO OPERATING PLANTS e
EFFECT OF ENVIRONMENT ON FATIGUE PROPERTIES.
- LOW AND HIGH CYCLE
- INITIATION AND PROPAGATION e WEAR RESISTANT ALLOYS WITH LOW COBALT.
e WELDING TECHNIQUES.
e NON DESTRUCTIVE TEST TECHNIQUES.
FOCUS ON INCREASED AVAILABILITY e
e COMPREHENSIVE UTILITY COOPEPATION EXCELLENT l
e 7/10/79 R.L.COWAN
OUTSIDE FUNDED PROGRA.MS IN PLANT MATERIALS AREA ELECTROCHEMICAL METHOD FOR DETECTION OF DEGPEE O NRC e
SENSITIZATION IN STAINLESS STEEL.
PARAMETRIC PIPING STUDIES OF STAINLESS EPRI e
QUALIFICATION OF NEAR TERM PIPING REMEDIES e
ALTERNATE PIPING MATERIAL QUALIFICATION PROGRA e
CARBON STEEL FATIGUE e
LARGE DIAMETER PIPE SCC MARGIN e
PIPING RESIDUAL STRESS IMPROVEMENT e
DOE (AS SUBCONTRACTOR) e ALTERNATE BWR WATER CHEMISTRY HELPS FOCUS DEVELOPMENT WORK ON COMMON OBJECTIVES ACCELERATES APPLICATION m
7/10/79 R.L.C0WAN
o,. -
FINAL PIPE TEST RESULTS - SCREENING PIPES MAXIMUM FACTOR PlPES F AILED BY HOUR 3 OFIGSCC ALLOY TESTED IGSCC ACCUMULATED IMPROVEMENT 1
7800*
19 14 REFERENCE TYPE 304
>20 REFERENCE TYPE 316L/
0" 8500 11 NUCLEAR GRADE
>20 REFERENCE TYPE 304L/
7800 9
0
>20 NUCLEAR GRADE 0"*
6900 8
>20 REFERENCE TYPE 347 3
0 1900
>20 REFERENCE TYPE 316 7400 8
0 REFERENCE TYPE CF-3
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- MEAN TIME TO FAILURE FOR REFERENCE TYPE 3 OF
- TWO PIPES FAILED BY TRANSGRANULAR MODE W CARBON = 239 HOURS.
F
'"ONE PIPE F AILED BY TRANSGRANULAR MODE WIT INTERGRANULAR FEATURES.
INTERGRANULAR FEATURES.
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SELECTION OF ALTERNATE ALLOYS FOR QUALIFICATION TYPE-316 NUCLEAR GR ADE TYPE 304 NUCLEAR GRADE CARBON <0.020%
- 0.06 - 0.16%
\\
' NITROGEN CURRENTLY RESTRICTED TO <0.10% BY ASME SA-240.
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SUMMARY
ALL CANDIDATE ALTERNATE ALLOYS EVALUATED (EXCEPT HIGH CARBON TYPE-316) WOULD BE MORE THAN ADEQUATE REPLACEMENTS FOR TY STAINLESS STEEL, BASED SOLEY ON IGSCC RESISTANCE.
e CONSIDERING ADVANTAGES AND DISADVANTAGES OF EACH CANDIDAT ALLOY, AS DEFINED AND WElGHTED BY THE DECISION ANALYSIS TYPE-316 NUCLEAR GR ADE W AS SELECTED AS THE BEST BALANCED CHOICE TO f REPLACE TYPE-304 IN THE BWR.
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COUNTERMEASURES FOR 304 SS PROGRAM STATUS AS OF 6/79 WELDS FAILED /
FACTOR OF MATERIALS CONDITIOM NO. HEATS WELDS TESTED IMPROVEMENI*
R FERENCE IYPE 304 SS 3
23/53 1
E (ALL HEATS) 3 0/33
~65 CORROSION RESISTANT CLAD PLuS SOLUTION HEAT IREATMENT (SHOP REMEDY) 2 0/22
~6 CORROSION RESISTANT CLAD AS DEPOSITED (FIELD REMEDY) 1 REFERENCE IYPE 304 SS 1
4/11 1
SENSITIZED MATERIAL 1
4/13 s6
, CORROSION RESISTANT CLAD DEPOSITED ON SENSITIZED MATERIAL l
'To DATE 1
3 j
COUNTERMEASURES FOR 304 SS PROGRAM STATUS AS OF 6/79 WELDS FAILED /
FACTOR OF MATERIALS CONDITION NO. HEATS WELDS TESTED IMPROVEMENT
- REFERENCE TYPE 304 SS 3
23/53 1
i f
(ALL HEATS)
J 3
0/33 s 65 f
Solution HEAT IREATMENT 15 s
1 2/24 HEAT SINK WELDING
.[*ToDATE e RECOMMENDED CO.UNTERMEASURES A.RE EFFECT i
- GENERIC RECIRCULATION PIPING RECOMMENDAT MADE TO ALL BWR PROJECTS
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