ML17309A548
| ML17309A548 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Point Beach, Arkansas Nuclear, Surry, Turkey Point, Crystal River, Ginna, Zion, Crane  |
| Issue date: | 06/21/1994 |
| From: | Hopkins J Office of Nuclear Reactor Regulation |
| To: | Hannon J Office of Nuclear Reactor Regulation |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9407010172 | |
| Download: ML17309A548 (82) | |
Text
~p,it RE0Iz P0 (a0
~,
ua R'+*<<>>
Docket Nos. 50-,244,-50-$ 50, 50-251, 50-266, 50-269 270st 50-,280, 50-281, 50-287, 50-289,-50-295, 50-301,'0-302, 50-304,-50-313,.
and 50-346 June 21, 1994 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555.0001 MEMORANDUM FOR:
John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation FROM:
SUBJECT:
Jon Hopkins, Sr. Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
SUMMARY
OF MEETING HELD ON HAY 19,
- 1994, WITH BABCOCK AND WILCOX REGARDING REACTOR VESSEL INTEGRITY On Hay 19,
A list of attendees is included as'nclosure 1.
The handout used at the meeting is included as Enclosure 2.
As shown in Enclosure 2, the items discussed during the meeting included reactor vessel integrity program, fracture mechanics methodology, bounding embrittlement trends, microstructural
- studies, and NRC comments.
By letter dated May 23,
- 1994, 88W confirmed that the information contained in Enclosure 2 may be considered non-proprietary.
Babcock and Wilcox said that it would provide a
common response to Generic Letter 92-01.
After the meeting, B&W agreed to provide this response 30 days subsequent (circa June 30, 1994) to the last response letter from the NRC to licensees (circa Hay 30, 1994).
The NRC staff said it would issue a
NUREG related to Generic Letter 92-01 data in July 1994.
B&W proposed a new method to determine the RT, value of Linde 80 welds relative to the pressurized thermal shock (PTF) screening criteria in the PTS rule.
The proposed method is a proactive initiative to address plant life extensions.
The new method would apply to all Linde 80 welds fabricated by B&W.
In the proposed method, all B&W fabricated Linde 80 welds would have an unirradiated reference temperature of -10 'F and a standard deviation of the unirradiated reference temperature of 0 F.
The standard deviation of the adjusted reference temperature would be 14 'F when credible surveillance data exists and 25 F when no surveillance data exists.
f 9407010172 940621 PDR ADOCK 05000244
~ gg ggSh
77 7
7 Hr. John N. Hannon Data previously provided by the B&WOG reported that B&W fabricate'd Linde 80 welds have an unirradi ated reference temperature of -5 'F and a standard deviation of unirradiated reference temperature 'ofi 17 'F.
RG 1.99; Revision 2
reports that the standard deviation for the adjusted reference temperature is 14 'F when credible surveillance data exists and 28 'F when no surveillance data exists.
The standard deviation in the unirradiated reference temperature and the adjusted reference temperature are needed to cover uncertainties in the values of the unirradiated reference temperature, copper and nickel
- contents, neutron fluences and calculational procedures.
The NRC staff indicated that the B&WOG must provide fracture toughness data to address the uncertainties and to demonstrate that the proposed new methodology is applicable to all B&W fabricated Linde 80 welds.
At the end of the meeting, the NRC staff and Babcock and Wilcox agreed to continue to communicate on this issue.
Enclosures:
1.
List of Attendees 2.
Meeting Handouts cc:
See next page Original Signed By!
3, 3, Hopkins Jon B. Hopkins, Sr. Project Hanager Project Directorate III-3 Division of Reactor Project III/IV Office of Nuclear Reactor Regulation TA:
I
-3 PDI II-3 PDIII-3 PDI EHC D:PD I 3 H.
Ru G. West J.
Ho kins D. HcDonald J
r snider J.
Hannon YES NO
(/7 94 YES NO
/
94 YES NO 3
94 YES NO m 94 YES NO L 94 YES NO Name:
G:(BWOG.NTG I
7
%7 77
l4 0
I I
1 fl I
t r
I i
t I
I
'f
Entergy Operations, Inc.
Arkansas Nuclear One, Unit I cc:
Hr. Harry M. Keiser, Executive Vice President 8 Chief Operating Officer Entergy Operations, Inc.
P. 0. Box 31995
- Jackson, Hississi ppi 39286 Mr. Charles B. Brinkman, Manager Mashington Nuclear Operations ABB Combustion Engineering Nuclear Power l2300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Hr. Nicholas S. Reynolds Minston 5 Strawn 1400 L Street, N.M.
Mashington, D.C.
20005-3502 Hr. Robert B. Borsum Licensing Representative BEW Nucleal Technologies 1700 Rockville Pike, Suite 525 Rockville, Naryland 20852 Senior Resident Inspector V.S. Nuclear Regulatory Commission P. 0.
Box 310 London, Arkansas 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 6]l Ryan Plaza Drive, Suite 1000 Arlington, Texas 7601l Honorable C. Doug Luningham County Judge of Pope County Pope County Courthouse Russellville, Arkansas 72801 Hs. Greta Dicus, Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 Mest Harkham Street Little Rock, Arkansas 72205-3867 Mr. Jerrold G.
Dewease Vice President, Operations Support Entergy Operations, Inc.
P. O. Box 31995
- Jackson, Mississippi 392&6 Mr. Robert B. McGehee Mise, Carter, Child I Caraway P.
O. Box 651
- Jackson, Mississippi 39286 Admiral Kinnaird R. McKee, USN (Ret) 214 South Norris Street Oxford, Maryland 21654 Nr. Jerry M. Yelverton Vice President, Operations ANO Entergy Operations, Inc.
Route 3 Box 1376 Russellville, Arkansas 72801
Florida Power Corporation CC:
Hr. Gerald A. Williams Corporate Counsel Florida Power Corporation NC-ASA P. 0.
Box 14042 St. Petersburg, Florida 33733 Nr. Bruce J. Hickle, Director Nuclear Plant Operations (NA2C)
Florida Power Corporation Crystal River Energy Complex 15760 W. Power Line Street Crystal River, Florida 34428-6708 Nr. Robert B. Borsum BN Nuclear Technologies 1700 Rockville Pike, Suite 525 Rockville, maryland 20852 Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Hr. Bill Passetti Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd.,
Tallahassee, Florida 32399-0700 Attorney General Department of Legal Affairs The Capitol Tallahaseee, Florida 32304 Hr. Percy N. Beard, Jr.
Sr. Vice President Nuc'lear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing (NA2I}
Crystal River Energy Complex 15760 W Power Line Street Crystal River, Florida 34428-6708 Crystal River Unit No.3 Generating Plant Hr. Joe Hyers, Directoi Oiv. of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Chairman Board of County Commissioners Citrus County 110 North Apopka Avenue Inverness, Florida 32650 Hr. Rolf C. Widell, Director Nuclear Operations Site Support (NA2I)
Florida Power Corporation Crystal River Energy Complex 15760 W Power Line Street Crystal River, Florida 34428-6708 Senior Resident Inspector Crystal River Unit 3 U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, Florida 34428 Hr. Gary Boldt Vice President - Nuclear Production (SA2C)
Florida Power Corporation Crystal River Energy Complex.
15760 W Power Line Street Crystal River, Florida 34428-6708
Mr. Donald C. Shelton Toledo Edison Company Davis-Besse Nuclear Power Stat'on Unit No.
1 CC:
Mary E. O'Reilly Centerior Energy Corporation 300 Madison Avenue
- Toledo, Ohio 43652 Mr. Milliam T. O'onnor, Jr.
Manager Regulatory Affairs Toledo Edison Company Davis-Besse Nuclear Power Station 5501 North State - Route 2
Oak Harbor, Ohio 43449 Gerald Charnoff, Esq.
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
W.
Washington, D.
C.
20037 Regional Administrator, Region III U. 5. Nuclear Regulatory Commission 801 Marrenville Road Lisle, Illinois 60532-4351 Mr. Robert B. Borsum Babcock E Wilcox Nuclear Power Generation Division 1700 Rockville Pike,- Suite 525 Rockville, Maryland 20852 Resident Inspector U. S. Nuclear Regulatory Commission 5503 N. State Route 2
Oak Harbor, Ohio 43449 Hr. John K. Mood, Plant Manager Toledo Edison Company Davis-Besse Nuclear Power Station 5501 North State Route 2
Oak Harbor, Ohio 43449 Robert E. Owen, Chief Bureau of Radiological Health Services Ohio Department of Health..
Post Office Box 118
- Columbus, Ohio 43266-0118
'ttorney General
, Department of Attorney General 30 East Broad Street
- Columbus, Ohio 43216 Mr. James M. Harris, Director Division of Power Generation Ohio Department of Industrial Regulations P. 0.
Box 825
- Columbus, Ohio 43216 Ohio Environmental Protection Agency DERR Compliance Unit ATTN:
Zack A. Clayton P. 0.
Box 1049
- Columbus, Ohio 43266-0573 Mr. James R. Williams State Liaison to the NRC Adjutant General's Department Office of Emergency Management Agency 2825 West Granville Road
- Columbus, Ohio 43235-2712
Dr. Robert C. Mecredy R.E. Ginna Nuclear Power Plant CC:
Thomas A. Hoslak, Senior Resident Inspector R.E.
Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road
- Ontario, New York 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs.
Donna Ross Division of Policy Analysis
& Planning New York State Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Charlie Donaldson, Esq.
Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Nicholas S.
Reynolds Winston
& Strawn 1400 L St.
N.W.
Washington, DC 20005-3502 Hs.
Thelma Wideman
- Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7370 Route 31
- Lyons, New York 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 West Fall
- Road, Room 11 Rochester, New York 14620
t Mr. J.
M. Haapton Duke Power Company Oconee Nuclear Station CC:
A. V. Carr, Esquire Duke Power Company 422 South Church Street Charlotte, North Carolina 28242-0001 J. Michael NcGarry, III, Esquire Minston and Strawn 1400 L Street, SM.
Mashington, DC 20005 Mr. Robert 8. Borsum Babcock 5 Milcox Nuclear Power Division Suite 525 1700 Rockville Pike Rockville, Haryland 20852
- Manager, LIS NUS Corp~ration 2650 McCoi'mick Drive, 3rd Floor Cl earwater, Fl orida 34619-1035 Senior Resident Inspector U. S. Nuclear Regulatory Coaeission Route 2, Box 610
- Seneca, South Carolina 29678 Regional Administrator, Region II U. S. Nuclear Regulatory Coaaission 101 Marietta Street, NW. Suite 2900 Atlanta, Georgia 30323 Max Batavia, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Supervisor of Oconee County Walhalla, South Carolina 29621 Mr. Steve Benesole Compliance Duke Power Company Oconee Nuclear Site P. 0.
Box 1439
- Seneca, South Carolina 29679 Mr. Marvin Sinkule, Chief Prospect Branch N3 U. S. Nuclear Regulatory Commissiea
)01 Marietta Street, NM. Suite 2900 Atlanta, Georgia 30323 Ms. Karen E. Long Assistant Attorney General North Carolina Department of Justice P. 0.
Box 629
- Raleigh, North Carolina 27602 Mr. G. A. Copp Licensing - EC050 Duke Power Company 526 South Church Street Charlotte, North Carolina 28242-086 Dayne H. Brown, Director Division of Radiation Protection North Carolina Department of Environment, Health and Natural Resources P. 0.
Box 27687
- Raleigh, North Carolina 27611-7687
8r. Robert E. Link Wisconsin Electric Power Company Point Beach Nuclear Plant Unit Nos. ) and 2
CC Ernest L. Blake, Jr.
- Shaw, Pittman, Potts 5 Trowbridge 2300 K Street, K.M.
Mashington, DC 20037 Hr. Gregory J. Maxfield, Hanager Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, bfisconsin 54241 Town Chairman Town of Two Creeks Route 3
Two Rivers, Misconsin 54241 Chairman Public Service Commission of Wisconsin Hills Farms State Office Building
- Nadison, Wisconsin 53702 Regional Administrator, Region III'.S.
Nuclear Regulatory Commission BOl Marrenville Road Lisle, I11inois 60532-4351 Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, Misconsin 54241
Virginia Electric and Power Company Surry Power Station CC:
Michael W. Maupin, Esq.
Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, Virginia 23219 Mr. Michael R. Kansler, Manager Surry Power Station Post Office Box 315 Sur ry, Virginia 23883 Senior Resident Inspector Surry Power Station U.S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Mr. Sherlock Holmes, Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr.
W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23209 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Robert B. Strobe, M.D., M.P.H.
State Health Commissioner Office of the Commissioner Virginia Department of Health P.O.
Box 2448 Richmond, Virginia 23218 Attorney General Supreme Court Building 101 North 8th Street Richmond, Virginia 23219 Mr. M. L. Bowling, Manager Nuclear Licensing
& Programs Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060 Mr. J.
P. O'Hanlon Senior Vice President - Nuclear Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060
Hr. T. Gary Broughton GPU Nuclear Corporation Three Hile Island Nuclear Station, Unit No.
1 CC:
Hichael Ross DEN Director, TMI-1 GPU Nuclear Corporation Post Office Box 480 Hiddletown, Pennsylvania 17057 John C. Fornicola.
Director, Licensing and Regulatory Affairs GPU Nuclear Corporation 100 Interpace Parkway Parsippany, New Jersey 07054 Jack S.
Metmore THI Licensing Hanager GPU Nuclear Corporation Post Office Box 480 Hiddletown, Pennsylvania 17057 Ernest L. Blake, Jr., Esquire Shaw, Pittman, Potts 8 Trowbridge 2300 N Street, NW.
Washington, DC 20037 Chairman Board of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, Pennsylvania 17120 Chairman Board of Supervisors of Londonderry Township R.D. 41, Geyers Church Road Hiddletown, Pennsylvania 17057 Hichele G. Evans Senior Resident Inspector (THI-1)
U.S. Nuclear Regulatory Commission Post Office Box 311 Hiddletown, Pennsylvania 17057 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Robert B. Borsum BEW Nuclear Technologies Suite 525 1700 Rockville Pike Rockville, Haryland 20852 William Dornsife, Acting Director Bureau of Radiation Protection Pennsylvania Department of Environmental Resources Post Office Box 2063 Harrisburg, Pennsylvania 17120
Florida Power and Light Company CC:
Harold F. Reis, Esquire Newman and Holtzinger, P.C.
1615 L Street, N.W.
Washington, DC 20036 Jack Shreve, Public Counsel Office of the Public Counsel c/o The Florida Legislature ill West Madison Avenue, Room 812 Tallahassee, Florida 32399-1400 John T. Butler, Esquire
- Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr. Thomas F. Plunkett, Site Vice President Turkey Point Nuclear Plant Florida Power and Light Company P.O.
Box 029100 Miami, Florida 33102 Joaquin Avino County Manager of Metropolitan Dade County ill NW 1st Street, 29th Floor Miami, Florida 33128 Senior Resident Inspector Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Commission P.O.
Box 1448 Homestead, Florida 33090 Mr. Bill Passetti Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Turkey Point Plant Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Center view Drive Tallahassee, Florida 32399-2100 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Plant Manager Turkey Point Nuclear Plant Florida Power and Light Company P.O.
Box 029100 Miami, Florida 33102 Mr.
H. N. Paduano, Manager Licensing L Special Projects Florida Power and Light Company P.O.
Box 14000 Juno
- Beach, Florida 33408-0420 Mr. J.
H. Goldberg President Nuclear Division Florida Power and Light Company P.O.
Box 14000 Juno
- Beach, Florida 33408-0420 Mr. Edward J.
Weinkam Licensing Manager Turkey Point Nuclear Plant P.O.
Box 4332 Princeton, Florida 33032-4332
Hr.
D. L. Farrar Commonwealth Edison Company CC:
Michael I. Hiller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60603 Dr. Cecil Lue-Hing-Director of Research and Development Metropolitan Sanitary District of Greater Chicago 100 East Erie Street Chicago, Illinois 60611 Phillip Steptoe, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60603 Mayor of Zion Zion, Illinois 60099 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 U.S. Nuclear Regulatory Commission Zion Resident Inspectors Office 105 Shiloh Blvd.
Zion, Illinois 60099 Regional Administrator U. S.
NRC, Region III 801 Warrenville Road Lisle, Illinois 60532-4351 Station Manager Zion Nuclear Power Station 101 Shiloh Blvd.
Zion, Illinois 60099-2797 Zion Nuclear Power Station Unit Nos.
1 and 2
ENCLO I
Attendees of Ma 19 1994 Meetin Between NRC Staff and Babcock and Wilcox NAME ORGANIZATION B.
K.
J.
E.
S.
K.
K.
L.
G.
W.
S.
D.
J.
D.
M.
J.
M.
S.
J.
W.
K.
G.
D.
Elliot Wichman Strosnider Hackett Collard Yoon Cozers Connor West Hazelton Sheng Miskiewicz Harrell Howell DeVan Gilreath Mitchell Katradis Taylor Pavinich Moore Lehmann McDonald NRC NRC NRC NRC Florida Power 8 Light BWNT NEI STS NRC Consultant NRC FPC Virginia Power BWNT BWNT Duke Power NRC NUS BWNT Grove Engineering BWNT GPUN NRC
, -Distribution:
', Docket File NRC L Local PDRs PDIII-3 R/F W. Russell/F. Miraglia 012G18 L. Reyes 012G18 J.
Roe 013E4 J. Zwolinski 013H24 N. Rushbrook 013E21 D. Pickett 013E21 J.
Hannon 013E21 OGC 015B18 E. Jordan 04D18 ACRS (10)
P315 W. Dean, EDO 17G21 B. HcCabe, EDO 17G21 EDO Region II Plants 017G21 EDO Region IV Plants 017G21 G.
- Greenman, Region III E. Herschoff, Region II R. Cooper, Region I B. Beach, Region IV A. Johnson 014D1 R. Crotean 014H22 A. Hansen 013E21 L. Wiens 014H25 B. Buckley 014H22 R. Hernan 014C7 C. Shiraki 013D9 L. Raghavan 014H15
- G.
Kalman 013H6
- R. Stransky 013E21
ENCLOSURE 0
l 9407010172
Ã:7c "
'. w,y pa x(s) gi(g+ggP v,QQf+@
~
g iWQ~4+P Vga E98 4
~, iP pe( P
~
+ ng q s'
j)'.
,/
~
g
~
0 s.~1I~+Lel54 Meeting with NRC Staff Rockville, Maryland
'ay 19, 1994 BBWNUCLEAR TECHNOLOGIES
REACTOR VESSEL INTEGRITY PROGRAM OVERVIEW FOR NRC hy THE B&W OWNERS GROUP REACTOR VESSEL WORKING GROUP MAY 19, 1994 AGENDA Section 1:00 pm ntroduction Meeting Purpose G. L. Lehmann/
K. R.
Wichman Reactor Vessel Integrity Program K., E. Moore 1
30 2:00 2:30 2:45 3:15 Fracture Mechanics Methodology Bounding Embrittlement Trends Break Microstructural Studies NRC Comments Program Critique NRC-Sponsored R&D Unresolved Issues K. K. Yoon G. L. Lehmann W. A. Pavinich NRC Staff 4
00 Adjourn
REACTOR VESSEL INTEGRITYPROGRAM Status ofIntegrated Surveillance Program o
Current tasks
~
Plan for future work
~f/JBBMIHUCLFAP I&MITECHNOLOG/ES
RVIP Cont.
RVWG Plan and Commitment
~
Obtain sufficient actual data and verified analytical methodology to demonstrate the safety of their reactor vessels
~
Minimize the effect of reactor vessel integrity issues on plant operation Duke Power Co Entergy Operations Florida Power Corp GPU Nuclear Corp Toledo Edison Co Commonwealth Ed Florida PAL Co Rochester GA,E Virginia Power Wisconsin EP Co Oconee-1, 2, 3 ANO-1 Crystal River-3 TMI-1 Davis-Besse Zion-1, 2 Turkey Point-1, 2 R. E. Ginna Surry-1, 2 Point Beach-1, 2 ffIBBW NUCLEAR MITECHNOLOGIES
RVIP Cont.
Master Integrated Reactor Vessel Surveillance Program (MIRVP) Review and Update
~
8AW-1543, Rev. 4
~
-100 plant specific RVSP capsules
~
8 88zW 177-FA plants
~
9 W plants with 88zW fabricated reactor vessels
~
Test reactor irradiations
~
ORNL/HSST program
~
NRC/NRL program jf!BBWNLJCLEAR
%MTECHNOLOGIES
P
RVIP Cont.
MIRVP Review and Update (Cont.)
~
14 RVWG capsules in 3 power reactors
~
13 pertinent Linde 80 weld metals
~
Fluences from 0.6E19 to 3E19
~
2 capsules to be used for annealing study including reirradiation
~
Specimens include 1T compact tension
~
2 RVWG capsules were tested
~
2 RVWG capsules to be tested in 1994
~
Schedule extends through 2008
~
1994 RVWG capsules include one at 1.6E19 n/cm' 177-FA IS 48 EFPY
~
W T/4 32 EFPY
~fJJBBHINVCLEAR IQCMTECHNOLOGIES
400 Linde 80 Weld Metals
~ 300 I-
~~200 K
'U
+ 100 0
Qg ~
68, g0 I
0 n
0 I
00 0
RG1.99/R2Pos.1 RG1.99/R2Pos.2 B&WRVSP 0
0 Q
B&WOG Svp. RVSP 0
W RVSP 0
0 2E+19 3E+19 Fluence, n/cm 2
4E+19 5E+ l9
RVIP Cont.
Five-Year Plan: List of Tasks I.
PRESSURIZED THERMAL SHOCK (PTS)
(1) Fracture Mechanics Analytical Methods (2) Microstructure and Material Properties (3) Regulatory Response II. PRESSURE/TEMPERATURE OPERATING'IMITS (1) ASME Appendix G Methodology III. LOWUPPER-SHELF ENERGY ISSUE (LUSE)
(1) LUSE Regulatory Issues IV. FRACTURE TOUGHNESS TEST METHODS (1) Charpy-Size Specimens
/SIBBWNUCLFAR
%4J TECHNOLOGIES
RVIP Cont.
Five-Year Plan: List of Tasks Cont.
V. COMMUNICATION (1) Industry and Code Review (2) Communication with the NRC (3) Status/Information Exchange VI. INFORMATIONBASE (1) Fluence Tracking System (2) Beltline Material Data Base VII. MASTER INTEGRATED RV MATERIAL SURVEILLANCE PROGRAM (MIRVP)
(1) MIRVP Capsule Testing and Evaluation fffBBWNUCLEAR LMTECHNOLOGIES
0
RVIP Cont.
Current RVWG-sponsored efforts in anal ical technolo includes the following:
e Evaluation of and familiarization with the FAVOR code, which may become the basis of future PTS reassessment
~
Engineering application of current and advanced fracture mechanics methods
~
Alternative method for radiation ernbrittlernent indexing
~JfjBBMINUCLEAR I&CMTECHNOLOG/ES
RVIP Cont.
RVWG is developing bounding embrittlernent trend curves for the Linde 80 class of welds.
RVWG-sponsored microstructural studies aim to establish a meaningful basis for radiation-induced ernbrittlernent.
IllBBMINLICLEAR MITECHNOLDG/ES
RVIP Cont.
Summa
~
Irradiated materials data is being provided by the MIRVP
~
Fracture toughness information Complete range of reactor vessel neutron exposure To investigate annealing and reirradiation To investigate irradiation temperature effects
~
Optimized methods for fracture mechanics test and analysis are being developed
~
Application to reactor vessel integrity assessment Ill BBWNUCLEAR CMTECHNOLOGIES
SECTION 2
B&WOWNERS GROUP. ACTIVITIESIN FRACTURE MECHAZGCS METHDOLOGY K. K. Yoon BRW Nuclear Technologies at BAW Owners Group Reactor Vessel Working Group Meeting with NRC Rockville Maryland May 19, 1994
CURRENT MEYHDOLOGYUPGRADE
- 1. Appendix 6 Methodology Update Code Activities - Section XI Residual Stress Analysis for BAW Fab. Vessels Cladding Operation and PWH'I'ladding Model Flaw Size Reduction Thermal KIModel 2.
PTS Analysis Methods Review of FAVOR Code New InQuence Function New Appendix A Method
- 3. Fracture Toughness Update Cognizance of ORNL Activities
ADVANCEDFRACTURE MECHANICS TOPICS A. Cognizance of NRC Sponsored Advanced FM Methodolgy Devlopments
- 1. Two Parameter J Theory Technology Follow-up Application Efforts 2.
Shallow Crack/Biaxial Loading Tests
- 3. ORNL Weibul Modeling of Fracture Toughness Curve Draft 5, Proposed ASTM Test Practice for Fracture Toughness in the Transition Range B. Development of Application Methodologies - BAWOG I.
B8zWOG Linde 80 Weld Metal Shallow Crack Testing 2.
Dodds-Anderson Constraint Correction 3.
Modified Boundary Layer Analysis 4.
J-Q Analysis Using Linde 80 Weld Metal Properties 5.
Alternative Method for Determining Initial RTNDT for Linde 80 Weld Metal 6.
Charpy Size Specimen Fracture Toughness Testing in Transition Temperature Range 7.
Devlopment of Direct Indexing Method for Irradiation Embrittlement
CURRENT FRACTURE MECHANICS METHODOLOGY UPDATE
Computation of Residual Stresses due to Application of Cladding in Reactor Pressure Vessels III. Procedure (a) 2-D Elastic-Plastic Creep Finite Element Analysis Finite Elemenl Model Qad Base Analysis Type:
Element Type:
DOF per node:
I of Nodes:
Coupled Temperaturo4is placement ABAOUS ~lemenl CAXSM Snoded axisymmetrfc. reduced htegratfon so8d elemenl 7 h Qad Mafetfaf 20 in Base Material 3
141 I~ 0.1875 in.
87.0 in.
Stress Analysis Boundary Conditions 8.0 in.
Internal Pressure~
Unilorm AxialDeformatfon L End Cap Pressure Loading tttttttttttttttt ttt Heat Transfer Boundary Conditions Qad hsulated Thsfde surface Toutslde surface hitfaf Conditions Temperature Only
Computation of Residual Stresses due to Application of Cladding in Reactor Pressure Vessels III. Procedure (continued)
(b) Loading History (Temperature/Pressure) 3500 Pressure (psia)
Temperature (deg F) 1600 3000 2500 2000 Pressure (psia) 1500 1000 500 Iaaaamaearaa e eea Taeaamaaa F aaal Poaa Wekl Heel Teaageaaae
~
Teal
~ ~e'
~
~ ~
~ 1 I ~
~ ~
~ ~
~ ~
~ ~e'
~
~
~
~
~
~
~
~
~
~
~
~ I
~ ~
~
I I
1400 1200 1000 600 Temp (deg F) 600 400 200 0
20 40 60 60 100 120 Time (hours)
Step Time (hrs)
Temp (deg F)
Pressure (psi)
Heat Transfer B.C.s Inside Outside Surface Surface Procedure 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 0
6.5 10.5 14.5 19.5 27.5 43.5 54 70 72 73.5 75 76.5 80 91.5 105.3 116.8 250 250 550 550 70 70 1100 1100 70 70 125 125 70 70 550 550 70 0
0 0
0 0
0 0
0 0
0 3150 3150 0
0 2250 2250 0
ambient ambient ambient ambient ambient ambient HS HS HS ambient HS HS HS ambient HS HS HS HS HS HS HS HS ambient HS HS HS ambient ambient ambient ambient ambient insulated insulated insulated Pre-Heat Intermediate Heat Treatment Final Post Weld Heat Treatment Shop Hydrostatic Test
~I Normal Operational Cycle
Computation of Residual Stresses due to Application of Cladding in Reactor Pressure Vessels IV. Results (a) General Stress Distribution (Same Trend for Both Axial and Hoop Stresses)
Following Application of Cladding, Heat Treatment, and Initial Service 50 40 30 20 Hoop Stress (ksi) 0
-10
-20
-30 87 88 89 90 91 92 93 94 95 Radius (rn)
(o',~ ueterinination of Cladding Stress-Free Temperatures in RPV Following Application of Cladding, Heat Treatment, and Initial Service 40 30
rr cvoal Stress 455 deg F
e
- Hoop Stress 374 deg "
20 Stress (ksi) 10
-10 0
100 200 300 400 500 600 Temperature (deg F)
Computation of Residual Stresses due to Application of Cladding in Reactor Pressure Vessels V. Comparison and Conclusions This work was based on the same general approach as given in Ref. 1. For this reason, we are able to make a direct comparison between the two approaches as shown below.
BWNT Ref.
1 Axial Hoop Axial Hoop
'%ifference Axial Hoop Residual Stress (ksi) 30.91 Stress Free Temp 455 (deg F) 25.41 28.50 24.00 374
- 400
-7.8
-5.5 (1)
Excellent correlation is found between BWNT and Ref.
1 for cladding stresses and the residual tensile stress field in the base material. All residual stress values were within 8% of one another.
(2)
Likewise, the predicted stress free temperaures for the cladding were equally reason-s
~
able.
It is important to note that while the stress in the cladding relaxes as the temperature is increased, the residual tensile field in the base metal is unaffected.
(3)
The assumption used by most researchers that the reactor pressure vessel is stress free at the final post weld heat temperature temperture is non-conservative for shallow flaws as no residual tensile stress field exists in the base metal.
This is primarily due to the fact that these references do not account for application of the cladding which is where the residual base metal stress field is created.
Ref.
1 B.R. Ganta, et. al., Cladding Stresses in a Pressurized 8'ater Reactor Vessel Folloiving Application ofthe Stainless Steel Cladding, Heat Treatment and Initial Service, ASME, PVP-Vol. 213, Pressure Vessel Integrity, 1991.
A REVIEW OF FLAW SIZE FACTOR BETWEEN SECTIONS III AND XI OF ASME BOILER AND PRESSURE VESSEL CODE K. K. Yoon.
Engineering and ProJect Services Division B&W Nuclear Technologies Lynchburg, Virginia INTRODUCTION This paper reopens the question of the appropriate safety factor to be applied to flaw size in Section XI, ASME Boiler Sc Pressure Vessel Code.
This is intended to promote discussion on the appropriaie safety factor and is not a proposal for any u
changes in the flaw acceptance standards of Section XI.
When thc flaw acceptance standards (IWB-3500) for Section XI of ASME Cod were developed, the acceptable flaw depth was selected as one-tenth of the Appendix G reference flaw"',
v;hich is a semielliptical surface flaw with a depth of one-fourth of the wall thickness and a length of six times the depth. The acceptable flaw depth for a planar flav'ith an aspect ratio equal 0.15 is then one-fortieth of the vessel wall thickness (IWB-3510-1 for wall thickness of 4-12 inches).
The logic behind this selection was documented in Reference 2 by R. R.
Maccary and it has remained the accepted basis.
The acceptable flav: depth is approximately 0.2 inch for an eight inch-thick reactor vessel; and this flaw was close to the limitof flaw-sizing capabiliiy of NDE methodology at the time.
It has been more than 22 years since the first flaw evaluation procedure for pressure vessels was developed by PVRC Committee on Toughness Requirements and was published as WRC Bulletin 175 to become the basis for Appendix G of Section GI. Recently, the Working Group on Operating Plant Criteria of Section XI, ASME Code, conducted a background study of Appendix G requirements regarding reactor vessel integrity"'.
When the Appendix G reference flaw size of a quarter of the wall thickness was determined, it was based on NDE techniques circa 1967.
Since then there have been vety significant improvements in NDE techniques, and many in-service inspections (ISI) of nuclear pressure vessels have been performed. The current ISI methodology can effectively detect and size flaws that are much smaller than thc two-inch reference flaw postulated in WRC-175. The Working Group on Operating Plant Criteria ofSection XIis revisiting the definition of the Appendix G reference flaw size.
A task group was formed to consider the latest NDE capabilities with fracture mechanics analysts to formulate a set of recommendations leading to a reduction in the postulated Appendix G reference flaw size.
The first obstacle this group encountered is that if the flaw size ratio of 10 is applied to a reduced Appendix G flaw size, the acceptable flaw is proportionately reduced and becomes unreasonably small.
This prompted a review of the basis document to determine the original reasoning for the establishment of the flaw size ratio of 10.
As a result of this revie~,
an oversight was found, which if rectified, greatly reduces the flaw size ratio. This is discussed below.
REVIEW OF BASIC PREMISE OF SECTION XI ACCEPTANCE STANDARDS Princi I Safety Criteria In the development of the acceptance standings, the guiding principles and rationale were derived from the following stated principal safety criteria'~
(a)
Tlute uRLy margins with respect to the stlU lulal integrityofthe components containing flaws within the limitsofthe 'allowable indication standards'. should not reduce the margins applied in the design of the component as related to the material's ductile behavior under conditions of normal plant operation.
(b)
"The safety margins applied in determining the stress intensity factors of materials and welds containing flaws within the limits of the 'allowable indication standards'hould be comparable to the margins of (a) above but related to the nonductilc behavior and material fracture toughness requirements specified in K. K. Yoon
The 26th National Symposium on Fracture Mechanics ROUND COMPACT SPECIMEN TEST METHOD FOR DETERMININGJ-R CURVES AND VALIDATIONBY J-TESTS K K. Yoon', L B. Gross', C. S. Wade'nd W. A.
VanDerSluys'BSTRACT:
A fracture toughness test method using a standard round compact tension (RCT) specimen is presented. This procedure is completely analogous to ASTM E 1152-87 standard for determining J-R curves using rectangular compact tension specimens. A slightly different round compact tension specimen design (BWRCT) is used by B&WOwners Group in their Integrated Reactor Vessel Material Surveillance Program"'IRVSP).
This specimen is analyzed by a finite element method to investigate whether the standard RCT compliance relationship is appropriate for use.
Validation tests using both square C(T) and RCT specimens were performed and the resulting J-R curves are compared.
It is concluded that using this procedure both square C(T) and BWRCT specimens yield similar J-R curves.
The ASTM standard test method for determining J-R curves was revised in 1987 and the new standard was issued as E 1152-87. This standard is for testing rectangular compact tension and bend specimens.
Round compact tension (RCT) specimens have been used for determining J-R curves for many years.
However, E 1152 does not include RCTs. The only reference in the ASTM standards relevant to RCT is a stress intensity factor equation found in ASTM E 399-83. Futato'"-'wrote a test procedure for RCTs in Babcock & Wilcox in 1984 based on the worL-. of Newman"'nd Underwood"'. A validation test was conducted to
'Advisory Engineers, B&WNuclear Technologies
-Section Manager and Scientist, Babcock & Wilcox Alliance Research Center
demonstrate that RCT specimen testing produces closely comparable J-R curves to those from standard C(T) testing in 1993.
This paper presents (1) a test procedure for round compact tension specimen testing in the same format of E 1152-87, (2) a finite element analysis of a round compact specimen to'determine compliance of a slightly different round compact tension specimen used in the B&WOwners Group IRVSP, and (3) the results of a validation testing to compare round C(T) with standard square C(T) specimens for identical weld metal and the data analysis by the proposed procedure and by current E-1152.
The resulting J-R curves are compared.
REFERENCES l.
A. L. Lowe, Jr., K. E. Moore, and J. D. Aadland, "Integrated Reactor Vessel Materials Surveillance Program for Babcock & Wilcox 177-FA Plants," ASTM STP-870, American Society of Testing Materials, Philadelphia, PA, 1985, p.931-950.
2.
R. J. Futato, "Round Compact Fracture Specimen Calibration," RDD:84:2839-0l-01:01, The Babcock
& Wilcox Company, Research
& Development
- Division, Alliance, Ohio, March 13, 1984.
3.
J. C. Newman, Jr., "Stress Intensity Factors'and Crack Opening Displacements for Round Compact Specimens,"
International Journal of Fracture, Vol. I7, No. 6,
~
December 1981, pp 567-578.
4.
J. H. Underwood, "K and Displacement for Disc Shaped Specimen," Unpublished data, U.S. Army Armament RD&E Center, Watervliet, New York.
ORNL DWG 93-11678 200
+
150 f 100 50 I
OO LO HSSI Weld 72W Irrad. at 288'C 2
Fluences n/cm
(>
1 MeY) o Weld Embrittled, 1.9E19
~
Duplex 1.56E19 Both spec. types. ASM'. To~12'C o /
Wdtt-Eeb. only, To ~ 14 C oo ro
-100
-50 0
50 100 Test Temperature
('C) 150 Figure S.
Crack-arrest toughness, K for irradiated HSSI weld 72W showing the results of both weldwmbrittled and duplex-type specimens.
250 a
200
. ~ 150 100 L
~( '50 0
O 0
ORNL-DWG 93-11679 v Unirradiated v Irradiated
~
'V','V
,'0 V
~g
~
~
~ ~
~
~
~
~
~
~
~
~....T.
84'C Y
~
T Y
r.'V'rrodioted 288oC 1.5 to 1.9E19 (>
1 IAeV)
HSSI WELD 72W (0.23% Cu)
Mean Fits to data 0
50 100 Test Temperature
('C) 150 Figure 6, Crack arrest toughness values for both unirradiated and irradiated 72W weld metal and for weldwmbrittled and duplex-type specimens.
HSSI WELD 72W DATA IRRADIATEDDATARESULTS O Fluence Range Measured Shift RG199 Predicted Shift Additional Margin 1.6 to 1.9K+19 151F (84C) 192F to 199F 41F to 4SF
ADVANCEDFRACTURE MECHANICS METHODOLOGY
Biaxial
~
~
500 400 300 H
33 x
200 a
W S
(mm) o 50 100 50
~
10 100 50 14 100 50 o
50 100 100 a
10 100 100 50 100 150 10 100 150 8
Biaxial Crucifrxm
~
Uniaxial Cruciform
~
0 ORNLAWG93-2782 ETD Sballaw Cock Deep Crack l
3
Tsbltt =BC
/I Uniaxial Cruciform(1)
(
Biaxial Cruciform (3) 100
~150 0100
-50 0
tt ()
50 100 Figure 28 Biaxial and unhxlal shallow-crack toughness data as function ofnormalized temperature 250 p~ Uoraxral Cruciform
~ Biaxial Crucilorm (3) 150 100
'enotes SE NB specimen 0
50 10 30 Crack Dcptb (mm)
Flgu're 2.4 Unlaxlal and biaxial toughness data as function ofcrack depth at T-RTNDT= <<IO'C
Constraint 500 400 300 a
B (mm) o SO 100 50
~
10 100 50 14 100 50 o
50 100 100 m
10 100 100 50 100 150 10 100 ISO D ~ ep Crack Tah s $$ C Shaltor Crack
~h
~0 200 I
m
~
~
100
-150
-100
-50 50 100 T RTeor (C)
Figure 324 HSST shallow-crack fracture toughness results as function ofnormalized temperature T RTNDT ORHLZWCS34948 ETD I
T RT
=-10 to 2$C 250
y ~ 198.17 + -1.5911x R~ 0.73068
~t C%
I 150 50 T
r Shallow Crack Dale:
Aan
=
182 QP&m Sld. Devlallon
= 27.2
@Palm Deep@rack Data:
Mean s 117 ilPaTm Std.
Devlatlon i 20A MPaRm 10 20 30 40 50 60 Crack Depth, a
(mm)
Figure 3.25 Toughness data at T-RTNDT= -25 to -10'C as function ofcrack depth
I
~
~
I I
~
~
~
~ ~
~
~
RVWG SHALLOWCRACKDATAANALYSIS 300 250
~ u Test Temp.
~
OF Q
-soF 200 hC X
E 150
~ U Ou Ou
~gu
~ c
~ c
~ U-
~c Qc 100 QHEC Oc QcOc 50 Occ R
u denotes tests in which significant tearing occured c denotes test in which cleav ge occ red wit out sig ificant t aring 0.10 0.20 0.30 0.40 0.50 Crack Length (a/W) 0.60 Fig 6 Values of Kmax as a function of temperature and crack length
1200 WF-292 (0 deg F), B&WRLD Division SE(B), a/W=0.17, W=2'in., 8=1in.
T,~281,n 10,Jic 694lbs/in, E/g ~432 1000 800 J
obs/in) 600 400
rsJc Measured
~
Jo Corrected 200 0
g
~
0.000 0.010 0.020 0.030 0.040 0.050 0.060 0.070 Crack Extension (in)
'I200 WF-292 (0 deg F), BRW R&,D Division SE(B). a/W=0.52. W=2 in., B=1in.
T. = 281. n
- 10. Jic 155.93 Ibs/in, E/g = 432 I
1000 800 J
(ibs/in) 600 a
400 200
rrJc Measured a
Jo Corrected 0.000 0.005 0.010 0.015 0.020 0.025 Crack Extension (in)
1200 NF-292 {40 deg F), B8 W RS.D Division SE{B), a/W=0.17. W=2 inB 1in.
T, ~311. n" 10. Jic" 273 lbslin, E/g ~435 1000 800 J
Obslin) 600 400
~ Jc Measured
~
Jo Corrected 200 g
~ ~
0,000 0.005 0.010 0.015 0.020 0.025 Crack Extension {in) 700 WF-292 (-50 deg F), BS W R8 D Division SE(B), a/W=0.52, W=2 in.. 8=tin.
T. =187, n" 10, Jic'47!bs/in,E/g =435 600 500 400 (lbs/in) 300 200 100
~ Jc Measured a
Jo Corrected 0.000 0.001 0.002 0.003 0.004 0.005 0.006 0.007 0.006 Crack Extension {in)
The 26th National Symposium on Fracture Mechanics SHALLOWCRACK TESTING OF Mn-Mo-Ni/LINDE80 WELD METALS K. K. Yoon B&WNuclear Technologies Lynchburg, Virginia W. A. VanDerSluys C. S. Wade Babcock & Wilcox Alliance Research Center Alliance, Ohio Apparent enhancement of fracture toughness due to constraint effects in test specimens with shallow crack depths was reported in the literature on A36, A533 and other materials, compared to the traditional deep crack specimen test data. None of the materials tested is weld metal.
The constraint effect on the crack depth is primarily a geometric parameter.
However, it may depend on the crack tip plastic zone shapes and stress field changes that may sho'v sensitivity to material differences.
A set of three point bend specimens was fabricated from a Mn-Mo-Ni/Linde80 weld metal and tested at two different temperatures in transition temperature range i.e. 0 and -50 degrees F, to establish the shallow crack effect on weld metals of significance to the commercial power reactor vessel integrity issue.
In addition, current Dodds and Anderson constraint correction methods, both 2D with crack growth and 3D, were applied and the results are compared with the uncorrected data.
The significance of this material data information on nuclear power plant safety evaluations under pressurized thermal shock transients is discussed.
Application of Two-Parameter (S-Q) Fracture Mechanics to Reactor Pressure Vessels Subjected to Pressurized Thermal Shocks Dbjective The long term primary objective of these analyses is to evaluate crack-tip stress fields in teactor pressure vessels (RPV) throughout a pressurized thermal shock (PTS) transient using the two-parameter J-Q &acture mechanics approach and incorporate small-specimen fracture toughness data in &acture mechanics assessments of RPVs.
Steps for ir.corporation of this technology into a RPV assessment approach
'r o
Development of boundary layer small scale yielding finite element methodolo o
Development of full body finite element models of RPVs containing circumferentially and axially oriented cracks.
o Perform PTS transient using the full body finite element models so that the applied Q stress can be evaluated using J-Q theory.
I o
Apply the Dodds-Anderson Scaling Model to experimental fracture toughness data to determine J,(Q,Temperature).
Application of Two-Parameter (S-Q) Fracture Mechanics to Reactor Pressure Vessels Subjected to Pressurized Thermal Shocks (Continued)
Step 1: Development of Boundary Layer Small Scale Yielding Finite Element Methodology o
A boundary layer approach using the finite element code ABAQUS, asstlming a rate-independent, J2 {isotropic-hardening) incremental plasticity theow), is adopted in evaluating the reference small-strain SSY crack tip fields in this study.
o The basis for the development of the BLM SSY approach is based on NUIKG/CR-6132 which allows for direct comparison and validation of the BWIW approach.
o Two BWNT FEA models were created.
BWNT Blunt Crack Tip Incorporates a crack-tip region with an initial root radius at the tip of 10 times the outer radius of the mesh BWNT Point Crack Tip Imposes an elastic-plastic singularity at the crack tip allowing for less computational time with only small variations in stresses from the more refined mesh.
o The plane strain reference fields for the two BWNT models was compared with the ORNL NUREG/CR-6132 model determined from the boundary layer model are shown in the figure on the next page.
As can be seen, excellent correlation can be seen for all three models as all solutions are within 3% of one another.
From a computational point-of-view, it is recommended that the BW1W elastic-plastic singularity specified finite element model be utilized in future investigations as the solutions are extremely accurate with the least computational time.
Application of Two-Parameter (S-Q) Fracture Mechanics to Reactor Pressure Vessels Subjected to Pressurized Thermal Shocks (Continued)
Step 1: Development of Boundary Layer Small Scale Yielding Finite Element Methodology (Continued)
O99 / CTO 4
BWNT SnyAaay Spaatied SSY BWNTSunt Tp SSY fco/J 10 Comparison of boundary layer small scale yielding finite elment analysis solutions between ORNL NUREG/CR-6132 and BWNT for A 533 B Steel at T=46'
t Application of Two-Parameter (J-Q) Fracture Mechanics to Reactor Pressure Vessels Subjected to Pressurized Thermal Shocks (Continued)
Step 2: Development of Full Body Finite Element Models of RPVs Containing Circumferentially and AxiallyOriented Cracks FEA Mesh for Axiall Oriented Flaw in RPV
ALTERNATIVEMETHOD FOR RADIATI N EMBRITTLEMENTINDEXIN 1.
SAW-2202 WF-70 Submittal - NRC Approval Received 2.
Additional Linde 80 Weld Metal Qualilication 3,
A Topical Report for AllLinde 80 Weld Metals 4.
Development of Direct Indexing Method to Model Radiation Embrittlement 5.
Development of Charpy Size Specimen Test Methods Dynamic versus Static Tests Effect of Test Temperature ASTM Standardization Activity
i 4-400 300 w 200 O
hY FRACTURE TOUGHNESS - WF-7O STATlC DATA BWOG DATA Kjc(med)
Kjc(95%)
KIR (-27)
VfOG DATA KIC (-27)
III 1
IIIIII I
f00
-200
-'f00 0
TEMPERATURE, F 100 200
250 200 V) 150
~ MI
>~ 100 50 k
/r::
I//// t k
/
//
///
k'
/
8 r
I..'
WF-70 Fracture Toughness
- WOGData k
Kjc(med) -35F KIb 95% -35F Kjc(med) OF KIb 95% OF Kjc(med) -70F KIb 95% -70F 0
-200
-100 0
Temperature, deg F 100 200
SUlNIVIARY BWOG is actively participating in the methodology updates in Appendix G of Section XI of ASME Code and the PTS analysis methods.
The NRC sponsored advanced FM methodology developments have very promising prospect of alleviating the PTS concern.
BWOG is trying to develop application methodologies using these new concepts.
BWOG anticipates great improvement in radiation embrittlement modeling by taking direct fracture toughness measurement approach.
Testing irradiated Charpy specimens in surveillance programs may provide direct fracture toughness which in turn generates fluence specific fracture toughness curves.
This approach willeliminate the indirect method of using Charpy impact energy data.
BOUNDING EMBRITTLEMENT TREND CURVES FOR LINDE 80 CLASS OF WELDS
Objective:
Provide fixityfor evaluating the embrittlement status RTndt Bc RTpts of the BEcWOG reactor vessels made of Linde-80 welds.
Method:
1 -
Demonstrate the appropriateness of an Un-Irradiated RTndt of -10 deg.F with a concurrent sigma I of zero (Oj for all Linde-80 welds.
2 -
Demonstrate that use of the R.G. 1.99 R-2 chemistry factors, based on the
~. mean of the chemistry, with a sigma shift of 14 deg.F when credible surveillance data exits and a sigma shift of 25 deg.F when no surveillance data exits, in accordance with the R.G. will yield bounding RTndt 8c RTpts trend curves by which embrittlement status can be assesed.
5/18/94
BSWOG Data for WF-70 S. WF-209 Walda (including Midland Beltline) 160 120 80 40 EK 0
~
NOT
~ T50
~
RTodt cv n
e coo I.~
s s
0 s
Data Source 160 ~
120 80 40 I
40
~80 BRWOG Data for WF.70 & WF.209 Welds (including Midland Baltline)
Qo QTndt R
p 0
0 10 20 30 40 50 60 X
~ Occnoc 2 0 Occ
+ Tioc I 0 2Ion2 k BaWOG RSVP Q Baw ludocol I480
~ HSST 0 IOdocxl I X ludoool 2 X IrnroocI 2
+ IMoooI4 lodoooI S BAW2202.XLS 5111/94
70 60--
50 -.
T50 ao
+ n-
~
40
) 30-CJ 20 aom1-a -
~
o a
Lee 10 -:tr.'+
0 Tndt BRWOG Data for WF-70 & WF-209 Welds (including Midland Beltline) 8 Oconee 2 0
Oconee 3
+
Zion 1 Zion 2 A
B8(WOO RSVP B&WMrdiend NBD 0
HSST 0
Midhnd 1 X
Mirhnd2 Midlend 3 Midlend 4
-80
-40 0
40 80 120 160 Temperature tdeg.F)
Midlands Cv Curve BAW2202.XLS 5/12IS4
B8WOG WF-70 8 WF-209 Un-Irradiated Capsules t:0
~~
0 Vl JDC e cn>
1I UJ CL 6$
O 80 70 60 50 40 30 20
~o
-'-4 0
J
/h RTn
.hh
(>
Oconee Unit 2 0
Oconee Unit 3 Zion1 Zion 2 D.B. WF-70
--.h RTndt
~-"
NDT Cv Mean WF-70 8 WF-209
-10
-50 0
50 100 150 200 250 300 350 0
Test Temperature, deg.F Lower Bound Upper Bound CV20ALL.XLS 5/18/94
I
B&WOG WF-25 8 SA-1526 Un-Irradiated Capsules t0 0
lA J24 e Cn -,i LQ ~
LLI Q.
C O
90 80 70 60 50 40 30 20 10 BTndt NDT 0-CD Surry t 0
Crystal River
+
TMI-1 0
TMI-2 Crystal River Oconee Unit 3
~ RTndt
~ NDT
- C v Mean Wf-25; SA-1526
-10
-50 0
50 100 150 200 250 300 350 0
Test Temperature, deg.F Lower Bound Upper Bound CV20AtLXLS 5/I8/84
BSWOG Un-Irradiated Charpy Data (excluding NF-70, NF-209, WF-25 5. SA1526) 90 80-70-60 50-40-CD
-100
-50 0
50 100 150 200
. 250 300 350 Test Temperature (deg.F) 0 W R ~
'W x-
+=xx Q xa "X
f, S
+g p
X3@ ~
Qe ~
% ~ & ~ H elx/
+=
X~
IP 10.~. y p gw~~~
AR01WF.193 0
R.S. WF.193
+
PB 2 WF.193 0
TP.3 SA 1101 TP4 SA.1094 D.B. WF 182
~
CRQ WF417 0
CR-3 SA.1686 X
D.B. WF 112
~
X D.B. SA.1136
+
PB 2SA 1283 Ginno SA 1038 Moon AI Cv Upper Bound Lower Bound CVUNIRR2.XLS 5/17/94
BEcWOG Un-Irradiated Charpy Data (excluding WF-70, WF-209, WF-25 5 SA1526) 90 80 70 50-40 CD
-100
-50 0
50:
100 150 200 Test Temperature (deg.F)
X 0:
X X
+ 'Og 0+ O~
~
fg X
+
r 4i'*
30
~
~
X)X X-- q 20 X aa X
0 AHO 1 WF.183 0
R.S.WF.I83
+
PB 2 WF.183 0
TP 3 SA 1101 TP4 SA.1084 D.B. WF.182 CRQ WF.87 0
CR4 SA 1686 X
D.B. WF 112 X
D.B. SA 1136
+
PB 2 SA 1283 Ginno SA 1038 Moon AIICv Upper Bound Lower Bound CVUNIRR2.XLS 5(16/94
B8 WOG Un-Irradiated Linde-80 Welds Charpy Curve Comparisons I:0 0
tO
< e cn -,j L
UJ CL L
05 0
90 80 70 60 50 40 30 20 10 0
RTn taa V
I'l-Ql NDt A AA E
IRT WF-25 & SA-1526 NDTWF-25 & SA-1526 IRT WF-70 &WF-209
"'~
NDTWF-70 &WF-209 C v Mean WF-25; SA-1526
-A Cv Mean WF-70 & WF-209
" " Cv Mean Other Linde80
-100
-50 0
50 100 150 200 250 Test Temperature, deg.F CV20ALL.XLS 5/18/94
Results:
Based on a review of the Un-Irradiated Charpy and IIIDTdata for the Linde-80 welds it is clear that the Charpy transition region for these welds willalways be above -10 deg.F. Thus, use of an upper bound un-irradiated RTndt of -10 deg.F is appropriate for the t.inde-80 welds.
NRCl.XLS 6/l8/04
I
WF-70 5. WF-209 Surveillance Data 250 200 150 CIZI-100 50 0
'/
/
I I
~
~i I
I
~
~
~
I
~
y
~
I I
I I
I.. l WI'-70 A 209 l(Tn<ll (50I'Il.bs) Doto Proposal IRT = 10; M = 2';
CI' 2I I R.Q. Pos. Nol II<'I'-5:M =
50;Cl'= 2l I 0
5E+ 18 1E <-19 1.5f + 19 2E+ 19 Capsule Fluence WELDPOS6.XLS 5/1 9/94
SA-1526 RWF-25 Surveillance Data 250-200-R'50-Cl
~e I
100 SA-1526 &WF-25 Data R.G. Pos. No. 2 R.G. Pos. No. I Proposed IRT = -IO; M = 28; CF =
223.6 50 0
5E+18 1E+19 Capsule Fluence 1.5E+19' 2E+19 WELDPOS6.XLS 5/17/94
==
Conclusion:==
Based on a review and comparison of the proposed RTndt bounding trend curves to the surveillance capsule RTndt it is apparent that the proposed method willyield conservatively bounding embrittlement (RTndt &RTpts) trends by which the status of these reactor vessels can be assessed.
SEC1'ION 4
MICROSTRUCTURAL STUDIES
OBJECTIVES Provide Physical Evidence To Support Observed MechanicalTest Data Trends.
Saturation/Stabilization
.Effective Copper Flux Effects irradiation Temperature Effects Provides Credibility To Correlations
~
Material Selected
~, Samples Have Been Prepared
~ Test Laboratories Selected APFIM -
Matrix GU Content/Precipitate Composition SANS -
Precipitate Size and Spacing Dislocation Density
MATERIAL IRRADIATEDCONDITIONS CU WF 209-1 Unirradiated 2.5E18 1.3E19 1.6E19.
0.35 WF 182-1 Unirradiated 5.9E18 9.6E18 1.3E19 0.24 WF 447 Unirradiated 4.0E18 1.7E19 0.03
CRP = Copper-Rich Precipitates UIVlD = Unstable Matrix Damage SMD = Stable lVlatrix Damage NEvTRoN FI,UgNgE
6 OD
=
Change in yield strength due to defects Change in yield strength due to precipitation 8, CTy
=
Change in yield strength
120-153
-100
-53 0
(4C) 100 lR) 2OO 250 100 80 M
~ 4
- 100, 80 QJ a) P 0
I 02'2 2
a) Number aasoclatod with data point
!ndlcalw tl.z number of tests having the same value.
2.0 L3 E
LO 0.5 0
ldv 14)
- 120
@1M eu 60 40 z
0 lo 0
8 ~
OWE; o tJn i,"radiat~<
~ Irradiated at 550'F lp'.
50 x 10 n/cm
-200
-100 0
100 200 300 Z0 Temperature ('F)
P 80 60 4t~
40 PC)
-150
-100
-50 0
50 100 150 200 250 t~
Z t
100 vl 80 60 40 20 0
120 0 OSNKATGI
~ RUBATO CR'A fUXXXL44l x N alQ 2.0 L5 1.0 05 0
160 100 80 I~
60 40 o
0 ~
0
~
~
~
u'f 120 80 40 "200
-100 0
100 200 300 400 500 TEHPCRATURE (of)
~ hcrD Appears To Be Small Based On Results Of Low Cu Linde 80 Welds I Supports Saturation Observation