ML20071L515
| ML20071L515 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1994 |
| From: | Allen Hiser NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-1426, NUREG-1426-V02, NUREG-1426-V2, NUDOCS 9408030136 | |
| Download: ML20071L515 (39) | |
Text
N U R EG-1426 Vol. 2 l'
Compila~ ion 0:' Reports :from Researca Sup;3orted Jy the Materials Engineering Branc 3, Division 0:? Engineering 1991 - 1993
,p U.S. Nuclear Regulatory Commission Ol'fice for Nuclear Regulatory Research p nucoq
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I AVAILABILITY NOTICE Avadabihty of Reference Mater als Cited in NRC Publications l
Most documents cited in NRC publications will be avai!able from one of the following sources:
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The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555-0001 i
i 2.
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t Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investi-gation notices; Licensoo Event Reports; vendor reports and correspondence; Commission t
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I Documents available from the National Technicalinformation Service include NUREG series
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f I
n NUREG-1426 Vol.2 l
Compilation of Reports from Research Supported by the Materials Engineering Branch, Division of Engineering 1991 - 1993 Manuscript Completed: May 1994 Date Published: June 1994 Compiled by A. L Iliser Materials Engineering Branch i
Division of Engineering Omcc of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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1 ABSTRACT Since 1%5, the Materials Engineering Branch, Division This report provides as complete a listing as practical of of Engineering.of the Nuclear RegulatoryCommission's formal technical reports submitted to the NRC by the i
Office of N uclear llegulatory itescarch, and its predeces-investigators working on these research programs. This i
sors dating back to the Atomic Energy Commission listing includes topical, final and progress reports, and is
}
(AEC), has sponsored research programs concerning the segmentedby topic area. In many cases a report will cover integrity of the primary system pressure boundary oflight several topics (such as in the case of progress reports of water reactors. The components of concern in these multi-faceted programs), but is listed under only one top-i rescarch programs have included the reactor pressure ic. Therefore, in scarching for reports on a specific topic, vessel (RPV), steam generators, and the piping. These other related topic areas should be checked also. The j
research programs have covered a broad range of topics, separate volumes of this report cover the following peri-including fracture mechanics analysis ar.d experimental ods work for itPV and piping applications, inspection method development and qualification, and evaluation of irradi-Volume 1: 1 % 5 - 1990
)
ation effects to RPV stecls.
Volume 2: 1991 - 1993 i
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iii NUREG-1426 1
1 1
CONTENTS Page i
Abstract........
iii Acknowledgments.....................
vii introduction.....
1 Advanced Reactors...
2 An n caling......................
2 Carbon Stect Castings............
2 i
Co r r e la t ion s......................................................................
3 Dosimetry.
3 Environmentally. Assisted Cracking and Fatigue 4
)
Fracture: Analysis.
7 Fracture Mechanics: Experimental - Component Testing.........................
11 Fracturc Mechanics: Experimental - Standard Specimen Testing...................
14 NDE - Continuous Monitoring (Acoustic Emission).............................................
17 NDE - In. Service Inspection........
17 Piping.....
19 Pressure Vessel St ects........................
21 Radiation Embrittlement 22 Reactor Pressure Vessel Integrity Assessments..................................................
24 l
Risk-llased Inspection.......
25 Stainicss Steel Sensitization 26 Thermal Aging (Cast Stainless Stect)..........................................
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l 1
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1 i
I v
l i
l ACKNOWLEDGMENTS
'Jhc assistance of S. Davis in preparing this report is appreciated.
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vii NUREG-1426
INTRODUCTION Since 1965, the Materials Engineering Ilranch, Divi-1975-1981 hictallurgy and h1aterials Research sion of Engineering, of the Nuclear Regulatory Com-liranch, Division of Reactor Safety mission's Of fice of Nuclear Regulatory Research, and Research, U.S. Nuclear Regulatory its predecessors dating back to the Atomic Energy Commission Commission (AEC), has sp(msored research pro-grams concerning the integrity of the primary system 1981-1986 Matcrials Engineering Ilranch, pressure boundary of light water reactors. 'lhe com.
Division of Engineering'Ibchnology, ponents of concern in these research programs have U.S. Nuclear Regulatory Commission included the reactor pressure vessel (RPV), steam generators, and the piping.'Ihese rescarch programs 1986-1993 hiaterials Engineering !! ranch, have covered a broad range of topics, including frac.
Division of Engineering, U.S. Nuclear ture mechanics analysis and experimental work for Regulatory Commission R PV and piping applications, inspection method de-velopment and qualification, and evaluation of irradi.
This report provides as complete a listing as practical ation effects to RPV steels.
of formal technical reports submitted to the NRC by i
the investigators working on these research pro-The branch sponsoring these research programs has grams. This listing includes topical, final and progress had various names and affiliations over the years, reports, and is segmented by topic arca. In many cases including the following:
a report will cover several topics (such as in t he case of progress reports of multi-faceted programs), but is 1965-1973 Reactor Vessels Ilranch, Division of listed under only one topic. Therefore, in scarching Reactor Development and for reports on a specific topic, other related topic
,Ibchnology, U.S. Atomic Energy areas should be checked also.
Commission The separate volumes of this report cover the follow-1973-1975 hietallurgy and hiatcrials Research ing periods:
liranch, Division of Reactor Safety Research, U.S. Atomic Energy Volume 1: 1965 - 1990 Commission Volume 2: 1991 - 1993 1
Compilation of Repons-1991-1993 Advanced Reactors fied and brief scoping plans developed for resolv-ing these issues.
Huddleston, R. L. Oak Ridge National Laboratory,
" Pressure Vessel Safety Research for Advanced Annealing Reactors. Semiannual Progress Report for December 1991 - March 1992," N U REG /CR-5900, Vol.1, No.1.
Cole, N. M. and Fridcrichs.T., MPR Associates,Inc.,
U.S. Nuclear Regulatory Commission, Washington,
" Report on Annealing of the Novovoronezh Unit 3 D.C., November 1992.
Reactor Vessel in the USSR," NUREG/CR-5760, U.S. Nuclear Regulatory Commission, Washington,
'lhe NRC sponsored " Pressure Vessel Safety D.C., July 1991.
Research for Advanced Reactors" project wasini-A U.S. delegation attended the thermal anneal-tiated at Oak Ridge National Laboratory in ing oper tion of the Novovoronezh Unit 3 reactor December 1992.'lhe objective of the project is to vesselin the USSR to evaluate the Soviet reactor identify and assist NRC in resolving any outstand-vessel annealing technology and to determme its ing safety issues in the materials, fabrication, de-applicability to PWR reactors in the U.S. Opera-sign, or environmental effects areas for the tions observed and described m this report m-reactor pressure vessel for the advanced reactors clude reactor vessel sample cutting, preparat,ons i
such that these issues do not negatively impact for annealmg, mstallation of anneahng appara-the design certification process. 'lhe major output tus, rnd initial heatup of the reactor vessel. The from this project during the first year will be white ann ating peration witnessed has been devel-papers summarizing the unresolved issues with p d te a munne operaton and appears applica-recommended plans for their resolution. The ble to U S. PWRs. Key areas requirmg further project will address two advanced evolutionary, work to canfirm applicability to U.S. reactors are two advanced passive, and four advanced reactor discussed.
systems. During this first report period progress included: development of a detailed technical ap-proach to be followed in identifying and resolving Carbon Siecl CaSitngs the issues; identification of docum ents containing Nanstad, R. K., et al., Oak Ridge National Labora-mformation needed for the assessments and re-tory, " Effects of Nonstandard Heat Treatment Tem-questing tbcse through official channels;identifi-perat urcs on Tensile and Charpy impact Properties of cation of relevant design code and regulatory Carbon-Steel Casting Repair Welds" NUREG/
documents and acquiring those which are not lo-CR-5972, U.S. Nuclear Regulatory Commission, cally available; development of a listing of infor-Washington, D.C., April 1993.
mation needed from the reactor manufacturers m preparation for visiting the manufacturers t o hold Carbon steel castings are used for a number of discussions with their technical personnel; devel.
different components in nuclear power plants, opment of a preliminary outline for the white including valve bodies and bonnets. Components papers; and initiation of document reviews.
are often repaired by welding processes, and both welded components and the repair welds are sub.
Huddleston, R. L and Swindeman, R. W., Oak Ridge jected to a variety of postweld heat treatments National Laboratory, " Materials and Design llases (PWHT) with temperatures as high as 899 C 1ssues in ASME Code Case N-47," NUREG/
(1650*F), well above the normal 593 to 677'C CR-5955, U.S. Nuclear Regulatory Commission, (1100 to l250*F) temperature range.The temper-atures noted are above the Al transformation Washington, D.C., April 1993.
temperature for the materials used for these A preliminary evaluation of the design bases components. A test program was conducted to (principally ASME Code Case N-47) was con-investigate the potential effects of such "nonstan-ducted for design and operation of reactors at dard" PWil'It on m echanical properties of carbon elevated temperatures where the time-depend-steel casting welds. Four weldments were fabri-ent effects of creep, creep-fatigue, and creep cated, two each with the shielded-metal-arc ratcheting are significant. Areas where Code (SMA) and flux-cored-arc (FCA) processes, with rules or regulatory guides may be lacking or inad-a high-carbon and low-carbon filler metal in each equate to ensure the operation over the expected case. All four welds were sectioned and given life cycles for the next. generation advanced simulated PWifIs at temperatures from 621 to high-temperature reactor systems, with designs 899 C (1150 to 1650*F)in increments of 56*C to be certified by the U.S. Nuc! car Regulatory (100*F) and for times of 5,10,20, and 40 h at each Commission, have been identified as unresolved temperature. Hardness, tensile, and Charpy issues. Twenty-two unresolved issues were identi-V-notch (CVN) impact tests were conducted for NUREG-1426 2
Compilation of Reports-1991-1993 the as-welded and heat-treated conditions. Re-NUREG/CR-3320 Vol. 2. U.S. Nuclear Regulatory sults were plotted versus a time-temperature re.
Commission, Washington, D.C., July 1992.
lationship (tempering parameter) to enable a more direct comparison of the effects of the vari-The metallurgical irradiation experiment at the ous PWiiT conditions. IIcat treatments at 621 Oak Ridge Research Reactor Poolside Facility and 677 'C (1150 and 1250'F) gave results amena-(ORR-PSF) is one of the series of benchmark ble to prediction, and regression analyses are experiments m the framework of the Light Water presented for those conditions. Ileat treatments Reactor Pressure Vessel Surveillance Dosimetry at 732 to 899'C (1350 to 1650'F), however, re-Improvement Program (LWR-PV-SDlP). The sulted in substantial changes in mechanical prop-g 1 of this program is to test, agamst well-erties of these Sh1A and FCA welds, with the established benchmarks, the methodologies and changes not amenable to prediction and highly data bases that are used to predict the irradiation dependent on the weld metal.1Icat treatments in embrittlement and fracture toughness of pres-that temperature range should not be applied to sure vessel and support structure steels.The pre-these materials without prior qualification for the diction methodology mcludes procedures for intended use, neutron physics calculations, dosimetry and spec-trum adjustment methods, metallurgical tests, and damage correlations. The benchmark experi-Correlations ments se rve io validate, improve, and standardize these procedures.He results of this program are Eason, E. D., et al., Modeling & Computer Services, implement ed in a set of ASThi Standards on pres-
"Multivariable Modeling of Pressure Vessel and Pip-sure vessel surveillance procedures. Rese, in ing J-R Da ta," N UREG/CR-5729. U.S. Nuclear Reg-turn, may be used as guides for the nuclear indus-ulatory Commission, Washington, D.C., May 1991, try and for the Nuclear Regulatory Commission (NRC). lb serve as a benchmark, a very careful M ultivariable models were developed for predict-characterization of the ORR-PSF experiment is ing J-R curves from availab!c data, such as mat cri, necessary, both in terms of neutron flux-fluence al chemistry, radiation exposure, temperature, spectra and of metallurgical test results. Statisti-and Charpy V-notch energy. He present work cally determined uncertainties must be given in involved collection of public test data, application terms of variances and covariances to make com-of advanced pattern recognition tools, and cali, parisons between predictions and experimental bration of improved meltivariable models. Sepa.
results meaningful. Detailed descriptions of the rate models were fitted for different material PSF physics-dosimetry startup experiments and groups, including RPV welds, Linde 80 welds, their results are reported.
RPV base metals, piping welds, piping base met-als, and the combmed database. Ihree different Williams, M. L, Louisiana State University, et al.,
types of models were developed, involving differ-
"Thmsport Calculations of Neutron Transmission ent combinations of variables that might be avail-Through Steel Using ENDF/ll-V, Revised ENDF/
Il-V, and ENDF/Il-VI Iron Evaluations," NUREG/
able for applications: a Charpy model, a pretrradiation Charpy model, and a copper-flu-CR-5648, U.S. Nuclear Regulatory Commission, ence model. In general, the best results were ob-Washington, D.C., April 1991.
tained with the preirradiation Charpy model.He The ENDF/B-VI evaluated nuclear data file has copper-fluence model is recommended only if been recently released by the U.S. National Nu-Charpy data are unavailable, and then only for clear Data Center during 1990. Among the most Linde 80 welds. Relatively good fits were ob-cagerly awaited new cross-section evaluations in tained, capable of predicting the values of J for this data collection are those for the natural iron pressure vessel stcels to within standard deviation isotopes, due to their importance in nuclear sys-of 13-18% over the range of test data.nc models tems analysis and because the previous ENDF/11 were qualified for predictive purposes by demon-data (version V, which was released in 1979) are strating their ability to predict validation data not known to underestimate the transmission of fast used for fitting, neutrons through steel structures such as reactor pressure vessels and radiation shielding. In this i
paper, a comparison is made of results obtained DoSIITletry from neutron transport calculations performed with these two ENDF/II versions (V and VJ) of McElroy, W. N., et al., Pacific Northwest Laboratory, iron data as well as an intermediate, revised ver.
" LWR Pressure Vessel Surveillance Dosimetry Im-sion V evaluation that was proposed in 1986. by-provement Program. PSF Startup Experiments,"
eral different response parameters that are j
Compilation of Reports-1991-1993 sensitive to high energy neutrons are examined, than in CP material, which indicates that irradi-for a variety of geometrical configurations and ationinduced segregation of impurity elements source spectra. It is found that the two newer and depletion of alloying elements are interde-iron evaluations substantially increase the trans-pendent.
mission of high energy neutrons through steel components with an incident fission spectrum Chung,11. M., et al., Argonne National laboratory, source. Preliminary estimates indicate that the
" Environmentally Assisted Cracking in Light Water version VI iron evaluation will considerably im.
Reactors. Semiannual Report, October 1991 - March prove the agreement between calculations and 1992," NUREG/CR-4667. Vol.14, U.S. Nuclear Reg-experimental dosimeter measurements used in ulatory Commission, Washington, D.C., August 1992, light water reactor pressure vessel fluence analy-
'lhis report summarizes work performed by sis. The calculated leakage spectrum of I?-T fu' Argonne National Laboratory on fatigue and emi-sion neutrons from an iron sphere is also ronmentally assisted cracking in light water reactors improved for energies above 4 MeV, but large during the six months from October 1991 through discrepancies with the measured spectrum are March 1992.'Ibpics that have been investigated dur-still observed at lower energies.
ng this period include: (1) fatigue and stress corro-sion cracking of low-alloy steel used in piping and in Environnientally-Assisted Cracking steam generator and reactor pressure vessels:
(2)mdiation-mduced segregation and irradiation-and Fat.igue assisted SCC of '13pe 3m SS after accumulation of relatively high fluence; and (3) update of a crack Chung,11. M., et al., Argonne Nat.ional Laboratory, growth data base for austenitic and ferritic steels in Environmentally Assisted Cracking m Light Water high-tempemture water. Existing data on fatigue of Reactors. Semiannual Report, April-September low-alkyy steel in LWR emironments have been 1990," NUR EG/CR-4667, Vol.11, U.S. Nuclear Reg-reviewed. Based on fmeture-mechanics models and ulatory Commission, Washmgton, D.C., May 1991.
engineering judgement, interim fatigue design This report summarizes work performed by Ar-curves are being developed that are consistent with gonne National Laboratory on environmentally available fatigue-life data. Microchem cal and mi-f assisted cracking in light water reactors during the crostructural changes in high-and commercial.
six months from April 1990 to September 1990, purity 'Iype 3m SS specimens from control-blade absorber tubes and a control-blade sheath from Crack-growthsate (CGR) tests were performed i
on a composite A533-Gr. B/Inconel-182 speci-operating BWRs were studied by Auger electron men in which a stress corrosion crack in the In-spectroscopy and samnmg electron microscopy, c
slow-strain-rate-tensile tests were conducted on tr-conel-182 weld metal penetrated and grew into the A533-Gr. B steel. CGR tests were also con-mdiated specimens in air and m simulated HWR ducted on conventional (nonplated) and Ni-or wate[at 289'C. Cmck growth data on fracture-me-Au plated A533-Gr. H specimens. CGR data on chames specimens of austenitic and ferritic stects m the A533-Gr. B specimens were compared with simulated BWR water, developed in this program the fatigue crack reference curves in the ASME ver the past 8 years, are compiled into a data base Boiler and Pressure Vessel Code,Section XI. Ap-along with references that contain details of test pendix A. Iligh-and commercial-purity (HP)and methods, material compositions, metalk> graph,c m-i (CP), respectively, specimens of Epc 304 SS from formation, and compansons of data with predictions of Section XI of the ASME Code, HWR absorber-rod tubes, irradiated during ser-vice to fluence levels of 6 x 1020 - 2 x 1025 n/cm2 Chung, H. M., et al., Argonne National Laboratory, (E > 1 MeV)in two reactors, were examined by "Emironmentally Assisted Cracking in Light Water Auger electron spectroscopy to characterize irra' Reactors. Semiann ual Report, October 1992 - March diation-mduced grain boundary segregation and 1993," NUREG CR-4667 Vol.16, U.S. Nuclear Reg-depletion of alhiying and impurity elements, ulatory Commission, Washington, D.C., September which have been associated with arradiation-ass-3993~
isted stress corrosion cracking of the steel. Inter-granular fracture surfaces in high fluence CP This report summarizes work performed by material were characterized by relatively high lev-Argonne National Laboratory on fatigue and en-els of Si, P, and Ni segregation. Segregation of the vironmentally assisted cracking (EAC) in light impurity elements and intergranular failure in water reactors (LWRs) during the six months the HP material were negligible for a similar flu-from October 1992 through March 1993. Fatigue ence level. However, grain boundary depletion of and EAC of piping, pressure vessels, and core Cr was more significant in high-purity material components in LWRs are important concerns as l
Compilation of Reports-1991-1993 extended reactor lifetimes are envisaged. Tbpics imately 1.0 ppm of CuCl in deoxygenated water that have been investigated include (1) fatigue of on the SCC susceptibility of 7} pes 316NG and 347 low-alhy steel used in piping, steam generators, SS and A533-Gr. Il and A 106-Gr. Il ferritic and reactor pressure vessels, (2) EAC of cast steels was determined in constant-extension-rate
{
stainless steels (SSs), (3) radiation-induced segre-tensile (CERT) tests at 200C. The CERT results gation and irradiation-assisted stress corrosion indicated that the alternative SSs were consider-cracking of 7}pe 304 SS after accumulation of ably more resistant to SCC than is sensitized Type relatively high fluence, and (4) EAC of low-alloy 304 SS. 'll e low-alloy ferritic st ecis exhibited only steels. Fatigue tests were conducted on medium-ductile fracture in this envirenment.
sulfur-content A106-Gr. Il piping and A533-Gr.
11 pressure vessel stects m simulated PWR water Kassner, T. E, et al., Argonne National Laboratory, and in air. Additional crack growth data were
" Environmentally Assisted Cracking in Light Water Reactors. Semiannual Report, April-September obtamed on fracture-mechanics specimens of cast austenitic SSs m the as-received and thermally 1991,"NUREG/CR-4667,Vol.13 U.S.NuclearReg-l aged conditions and chromium-nickel-plated ulatory Commission, Washington, D.C., March 199' A533-G r. Il st cel in sim ulated boiling-water reac-This report summarizes work performed by tor (llWR) water at 289'C. The data were com.
Argonne National 12boratory on fatigue and emi-pared with predictions based on crack growth ronmentally assisted eracking in light water reactors correlations for ferritic st cels in oxygenated water during the six months from April 1991 through Sep-and correlations for wrought austenitic SS in oxy.
tember 1991. 'Iopics that have been investigated penated water developed at ANL and rates in air during this period inc'lude: (1) fatigue and stress from Section XI of the ASME Code. Microchem.
corrosion cracking (SCC) of low-alloy steel used in ical and microstructural changes in high and piping and in steam generator and reactor pressure commercial-purity 'I}pe 304 SS specimens from vessels; (2) role of chromate and sulfate in simu-control-blade absorber tubes and a control-blade lated boiling water reactor (llWR) water on SCC of sheath from operating IlWRs were studied by sensitized 'I}pe 3M SS; and (3) radiation-induced Auger electron spectroscopy and scanning elec-segregation (RIS) and irradiation-assisted SCC of tron microscopy.
Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-Kassner, T. E, et al., Argonne National Laboratory, content A533-Gr. B and A106-Gr. B steels in high.
" Environmentally Assisted Cracking in Light Water purity, (HP) deoxygenated water, in simulated Reactors. Semiannual Report, April-September pressurized water reactor (PWR) water, and in air.
1989," NUREG/CR-4667, Vol. 9. U.S. Nuclear Reg-Crack-growth-rates (CGRs) of composite speci-ulatory Commission. Washington, D.C., March 1991.
mens of A533-Gr. II/Inconel-182/Inconel-M (plated with nickel) and homogeneous specimens This report summarizes work performed by of A533-Gr. B were determined under small.
Argonne National Laboratory on environmental-amplitude cyclic loading in HP water with -300 ppb ly assisted cracking in light water reactors during dissolved oxygen. CGR tests on sensitized Type 304 the six months from April 1989 to September SS indicate that low chromate concentrations in 1989. 'Ibpics that were inv:stigated inctude BWR water (25-35 ppb) may actually have a benefi-(1) stress corrosion cracking (SCC) of A533-Gr.
cial effect on SCC if the sulfate concentmtion is B steel in simulated boiling water-reactor emi-below a critical level. Microchemical and micro-ronments (2) SCC of'l} pes 347 and CF-3 cast structural changes in HP and commercial-purity duplex stainless stcel(SS), and (3) effects of heat-7}pe 304 SS specimens from control-blade absorber to-heat variation on SCC of 7}pe 304 SS. Crack-tubes used in two operating IlWRs were studied by growth-rate (CGR) tests were performed on Auger electron spectroscopy and scanning c!cctron conventional (non-plated) and nickel-or gold-microscopy, and slow-strain mte-tensile tests were plated A533-Gr. B specimens to provide insight conducted on tubular specimens in air and in simu-into whet'icr the surface layer on the low-alloy lated BWR water at 289'C.
steel, either oxide corrosion products or a noble metal, influences the overall SCC process. CGR Maj.umdar, S. and Chopra, O. K Argonne National tests were also conducted on specirnens of'I}pe Laboratory,
- Interim Fatigue Design Curves for Car 347 SS with different heat-treatments, and a spec-bon, low-alk)y, and Austenitic Stainless Steels m,-
imen of CF-3 cast SS with a ferrite content of LWR Emironments," NUREG/CR-5999, U.S. Nu-15.6% CGR data oi, these specimens were com-clear Regulatory Commission, Washington, D.C.,
pared with reference fatigue crack growth curves
^Pfil1993' in the ASME Boiler and Pressure Vessel Code, Existing data in the literature on fatigue of car.
Section XI, Appendix A.'lhe influence of approx-bon, low-alkiy, and austenitic stainless steels in 5
Compilation of Reports-1991-1993 LWR environments are reviewed. It is found that Ruther, W. E., et al., Argonne National Laboratory, Imth temperature and dissolved-oxygen concen-
" Environmentally Assisted Cracking in Light Water tration in water significantly affect fatigue life. At Reactors. Semiannual Report, April-September the very :ow dissolved-oxygen !evcls charaetcris-1992," NUREG/CR-4667, Vol.15, U.S. Nuctcar Reg-tic of pressurized water reactors and boiling water ulatory Commission, Washington, D.C., June 1943.
reactors with hydrogen-water chemistry, environ-33;3 g
gg mental effects on fatigue life are modest. Ilowev-cr, at higher dissolved-oxygen les els (2 100 ppb),
Argonne National Laboratory on fatigue and en-significant reductions m fatigue life can occur, vironmentally assisted cracking (EAC) in light lhe susceptibility of carbon and low-alloy steels water reactors (LWRs) during the six months to reduced fatigue life is strongly related to sulfur from April 1992 to September 1992.1bpics that concentration. Although the fatigue hves of aus-have been investigated include: (1) fatigue and tenitic stainless steels may be reduced, the reduc-stress corrosion cracking (SCC) of low-alloy steel tions are much smaller than those observed in used in piping, steam generaters, and reactor high-sulfur carbon and low-alloy steels. In oxy.
pressure vessels;(2) EAC of cast stainless steels genated water, fatigue life depends strongly on (SSs); and (3) radiation-induced segregation and strain rate. Interim fatigue design curves are pro-irradiation-assisted SCC oflype 304 SS after ac-posed that take into account temperature, dis-cumulation of relatively high fluence. Data on soh ed-oxygen level in the water, the sulfur level fatigue of low-alloy steel in LWR environments have been reviewed. Ilased on fracture-m the steel, and stram rate. Design curves for cartmn and low-alloy steels for lives up to 10 mechanics models and engineering judgment, in-cycles are also proposed.
terim fatigue design curves were developed that are consistent with available fatigue-life data.
Crack growth data were obtained on fracture-me-chanics specimens of A533-Gr.11 and A106-Gr.
Ruther, W. E., et al., Argonne National Laboratory, 11 ferritic stects and on cast austenitic SSs in the
" Environmentally Assisted Cracking m Light Water as-received and thermally aged conditions in sim-Reactors. Semiannual Report, October 1989 - March ulated IlWR water at 289'C. The data were ccm-1990," N UR EG /CR-4667, Vol.10, U.S. Nuclear Reg-pared with predictions based on crack growth ulatory Commission, Washington, D.C., March 1991, correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxy.
This repon summarizes work performed by genated water developed at ANL and rates in air Argonne National Laboratory on environmental.
from Section XI of the ASME Code. Microchem-ly assisted cracking in light water reactors during ical and microstructural changes in high-and the six months from October 1989 to March 1990.
commercial-purity lype 304 SS specimens from Low-cycle fatigue tests were performed on Type control-blade absorber tubes and a control-blade 316NG SS to better understand the effects of sheath from operating IlWRs were studied by cyclic strain range, frequency, and temperature Auger electron spectroscopy and scanning elec-on fatigue life in air and in simulated IlWR water, tron microscopy. Slow-stram rate-tensde tests and to assess the degree of conservatism in the were conducted on irradiated specimens in air ASME Code Section III fatigue design curves.
and simulated BWR water.
Fracture mechanics crack-growth-rate tests were carriedoutonacompositespecimenof A533-Gr.
Shack, W. J., et al., Argonne National Laboratory, II/Inc(mel-182/Inconel-600, plated with mckel,
.* Environmentally Assisted Cracking in Light Water to establish whether a transgranular crack will Reactors. Semiannual Report, October 1990 - March imtlate m the ferritic steel from an intergranular 1991," NUR EG/CR-4667, Vol.12, U.S. Nuclear Reg-crack in the Inconel-182 w eld metal at low stress ulatory Commission, Washington, D.C., August 1991.
intensity associated with crack growth in the In-This report summarizes work performed by conel-182 weld metal. Irradiated Iainless steels Argonne National Lalmratory on environmental-from absorber-rod tubes, control-rW cladding, ly assisted cracking in light water reactors during and flux thimbles of several llWF6 and PWRS the six months from October 1990 to March 1991.
were obtained to investigate the nature and ex.
Fatigue life of A533-Gr. Il pressure vessel steel tent of radiation-induced segregation in the steels was st udied in high-purity (I1P) deoxygenated wa-and correlate it with susceptibility to intergranu-ter, in simulated PWR water, and in air. Fatigue lar failure in the materials. Specimens have been data are compared with the design curve in See-prepared for Auger electron spectroscopy analy-tion ill Appendix A of the ASME Iloiler and ses of segregation of alloying elements on inter-Pressure Vessel Code. Equations in Section XI of granular fracture surfaces.
the ASME Iloiter and Pressure Vessel Code that NUREG-1426 6
Compilation of Reports-1991-1993 l
relate crack growth rates (CGRs) of ferritic stecis plarie strain conditions, were applied to addition-to loading parameters have been modified to in-al large-scale data with the objective of validating I
corporate CGR data that we recently acquired at models in the planc stress-to-planc strain domain j
high load ratios. The effect of water flow rate on before applying them to positive out-of-plane the SCC behavior of 7)pe 316NG stainless steci strain conditions. The general finding was that (SS) was investigated in fracture-mechanics CGR applications of the models resulted in predictions tests in lip oxygenated water at 289'C. Corrosion of fracture behavior that conflicted with existing fatigue curves for austenitic SS in Section XI of experimental data censidered relevant to the the ASME Holler and Pressure Vessel Code have problem. llecause of the conflicting results, it is been modified to be more consistent with SCC apparent that testing of RPV stects is required:
data in simulated LWR cnvironments at high load (1) to determine the magnitude of out-of-plane ratios. High and commercial-purity (CP) speci-biaxial loading effects on fracture toughness; and mens of 'lype 304 SS from HWR absorber-rod (2) to provide a basis for development of predic-tubes, irradiated during service in two reactors to tive models.'Ihis course of action is necessary to fluence levels of 1.4-2 x 10a n/cm2(E > 1 McV),
support a refined treatment of in-plane and out-were examined by Auger clectron spectroscopy to of-plane constraint effects in FFS analysis. Pro-characterize irradiation induced grain boundary posed in this report are critcria for a biaxial speci-segregation and depletion of alloying and impuri-men that would form the basis of a testing ty elements, which have been associated with program designed to provide data to explain dif-irradiauon-assisted SCC of the stcel. Slow-strain-ferences between theoretical predictions and rate tensile tests have been conducted in air and measured material behavior. Results of design in simulated IlWF. water on specimens obtained studies on the biaxial specimen will be presented from the irradiated CP'Iype 304 SS absorber-rod in a future report from the Heavy-Section Stect tubes.
Technology Program.
Hass, H. R., et al., Oak Ridge National Laboratory, Fracture: Analysis "CSNI Project for Fracture Analyses of Large-Scale International Reference Experiments (Project FAL-Ilass, H. R., et al., Oak Ridge National Laboratory, SIRE)," NUREG/CR-5997. U.S. Nuclear Regula-
" Constraint Effects on Fracture 'Ibughness for Cir-tory Commission, Washington, D.C., June 1993.
cumferentially Oriented Cracks in Reactor Pressure Vessels," NUREG/CR-6008, U.S. Nuclear Regula-
'Ihis report summarizes the recently completed tory Commission, Washington, D.C., Aut,ust 1992.
Phase I of the Project for Fracture Analysis of large-scale International Reference Experiments Pressurized-thermal-shock (FTS) loading pro-(Project FAMIRE). Project FAMIRE was created duces biaxial stress fields in a reactor pressure by the Fracture Assessment Group (FAG)of Princi-vessel (RPV) wall with one of the principal pal Working Group No. 3 (PWG/3) of the Organi-stresses aligned parallel to postulated surface zation for Economic Cooperation and Develop-cracks in either longitudinal or circumferential ment (OECD)/ Nuclear Energy Agency's (NEA's) welds. The limited quantity of existing biaxial test Committee on the Safety of Nuc! car Installations data suggest a significant decrease of fracture (CSNI). Motivation for the project was derived toughness under out-of-planc (i.e., parallel to the-from recognition by the CSN1-PWG/3 that incon-crack front) biaxial loading conditions when com-sistencies were being reve:ded in predictive capabili-pared with toughness values obtained under uni-ties of a variety of fracture assessment methods, axial conditions. Any increase in crack-tip especially in ductile fmeture applications. As a con-constraint resulting from these out-of-plane biax-sequence, the CSNI/ FAG was formed to evaluate ial stresses would act in opposition to the in-planc fracture prediction capabilitics currently used in constraint relaxation that has been previously safety assessments of nuclear components. Mem-demonstrated for shallow cracks. Consequently, bers are from laboratories and research organiza-understanding of both in-plane and out-of-plane tions in Western Europe, Japan, and the United crack-tip constraint effects is necessary to a re-States of America (USA). On behalf of the CSNI/
fined analysis of fracture initiation from shallow FAG, the U.S. Nuclear Regulatory Commission's cracks under FTS tmnsient loading. This report is (NRC's) Heavy-Section Stcel 'Rchnology (HSST) the second in a series investigating the potential Program at the Oak Ridge National Lalmratory impact of far-field out-of-plane strer/+s and (ORNL) and the Gesclischaft fur Anlagen-und strains on fracture initiation toughness. Selecd Reaktorsicherheit (GRS), Koln, Federal Republic fracture prediction models, previously validated of Germany (FRG) had resp (msibility for for small-scale fracture specimens under nearly organization arrangements related to Project 7
Compilation of Reports-1991-1993 FALSIRE.1he group is chaired by 11. Schulz from the current state of fracture prediction technolo-GRS, Koln, FRG.
gy is reasonably well advanced, more work is needed to provide analysis methods Dodds R. ll., University ofIllinois, et al.,"Continu.
capable of accurately predicting ductile crack um and Micromechanics 'Ileatment of Constraint in extension.
Fracture " NUREG/CR-5971, U.S. Nuclear Regula-tory Commission, Washington, D.C., July 1993.
Keency-Walker, J., et al., Oak Ridge National Labo-1Wo complementary methodologics are descr. bed ratory, "An Investigation of Crack-Tip Stress Field Criteria for Predicting Cleavage-Crack Initiation,"
to quantify the effects of erack;tip stress tnaxiality NUREGICR-5651, U.S. Nuclear Regulatory Com-(constraint) on the macroscopic measures of clas-tic-plastic fracture toughness, J and Cf0D. In mission, Washington, D.C., September 1991.
the continuum mechanics methodology, two pa-CMavage crack initiation in large-scale wide-plate rameters, J and Q, suffice to characterize the full (wp) spc~!:n could not be accurately predicted range of near-tip environments at the onset of from small. compact (Cl) specimens utilizing a lin-fracture. A micromechanics methodology is de-carclasticfracturemechanics,K e, methodology in t
scribed which predicts the toughness locus using the wide-plate tests conducted by the Heasy-crack-tip stress fields and critical J-values from a Section Steel 1bchnology Program at Oak Ridge few fracture toughness tests. A robust microme-National Laboratory, crack initiation has consistent-chanics model for cicavage fracture has evolved ly occurred at stress intensity K values ranging i
from the observations of a strong, spatial self-from two to four times those predicted by the CT i
similarity of crack-tip principal stresses under in-specimens 'Ihe work centers around nonlinear two.
creased loading and across different fracture and three-dimensional finite-element analyses of specimens.This report explores the fundamental the crack-tip stress fields in these geometries. Anal-i concepts of the J.Q description of crack-tip fields, yses were conducted on CT and WP specimens for the fracture toughness locus and micromechanics which cleavage initiation fracture had been mea-approaches to predict the variability of macro-sured in laboratory tests. ~1he local crack-tip fields scopic fracture toughness with constraint under generated for these specimens were then used m elastic-plastic conditions. Computational results the evaluation of fracture correlation parameters to are presented for a surface cracked plate contain-augment the K parameter for predicting cleavage I
i ing a 6:1 semi-clliptical, a-t/4 flaw subject re-initiation. hirameters of hydrostatic constraint and i
mote uniaxial and biaxial tension.
of maximum principal stress, measured volumetri-cally, are included in these evaluatiorr lhe results IIiser, A. L and Mayfield, M. E., U.S. Nuclear Regu-suggest that the cleavage initiation process can be latory Commission," Proceedings of the Seminar on correlated with the hxal crack-tip fields via a maxi-(
Assessment of Fracture Prediction Technology; hp-mum principal stress criterion Insed on achiesing a i
ing and Pressure Vessels," NUREG/CP-0037, U.S.
cntical area within a critical stress contour,1his Nuclear Regulatory Commission, Washington, IAC.,
criterion has been successfully applied to mrrelate l
February 1991.
cleavage initiation in ZF--Cr and WP specimen i
geometries.
1he 1990 Pressure Vessel and Piping Conference, I
sponsored by the American Society of Mechani-Keency-Walker, J and 11 ass,11. R., Oak Ridge Na-cal Engineers (ASME), was held in Nashville, tional Laboratory,"A Comparison of Analysis Meth-Tennessee from June 18 to June 21,1990. As part odologies for Predicting Cleavage Arrest of a Deep
{
of that conference, representatives from the Crack in a Reactor Pressure Vessel Subjected to USNRC and AEA 'Ibchnology in the United Pressurized-lhermal Shock Loading Conditions,"
l Kingdom jointly organized two panel sessions to NUREG/CR-5793, U.S. Nuclear Regulatory Com-discuss the current state of fracture prediction mission, Washington, D.C., September 1992.
technologies for piping and pressure vessels. A total of nine presentations were given, contrast.
Several calculational procedures are compared ing analytical predictions with experimental re-for predicting cleavage arrest of a deep crack in sult, This document provides summaries of each the wall of a prototypical reactor pressure vessel presentation and copies of the pertinent figures (RPV) subjected to pressurized-thermal-shock and other visual aids. This information has been (l'rS) types of loading conditions.1hree procc-compiled and published to permit reasonably dures examined in this study used the following prompt disseminations of the information pres-models: (1) a static finite-element model (full ented. Ilased on the information presented dur-bending): (2) a radially constrained static model; ing these two panel sessions,it appears that.w hile and (3) a thermoclastic dynamic finite-element NUREG-1426 8
1 l
1 Compilation of Reports-1991-1993 I
model. A l'IS transient loading condition was Regulatory Commission, Washington, D.C., Novem-selected that produced a deep arrest of an axially ber 1992.
l oriented, initially shallow crack according to cal.
Licensing issues within the nuclear industry dic-culational results obtamed from the static (full-tate a need to investigate the effects of cladding bending) model. Results from the two static on the extension of small finite-length eracks near models were compared with those generated the inside surface of a vessel. Limited experimen-from the detailed thermoclastic dynamic finite tal data and analyses indicate that cladding can element analysis. Itc dynamic analyses modeled inhibit the propagation of certain shallow flaws.
cleavage-crack propagation using a nod-release
.Ihis report describes an analytical study which techmque and application-and generation-mode was carried out to determine (1) the minimum methodologies. Comparisons presented here m-flaw depth for crack initiation under l'rS loading dicate that the degrec to which dynamic solutions for semicircular surface flaws in a clad reactor can be approximated by static models is highly pressure vessel and (2) the impact, in terms of the dependent on several factors, meloding the mate-conditional probability of vessel failure, of usinga rial dynam,c fracture curves and the propensity i
semicircular surface flaw as the initial flaw. The for cleavage reimtiation of the arrested crack tm-analytical results indicate that, for initiation, a der 11S loading conditions. Additional work is much deeper critical crack depth is required for required to develop and salidate a satisfactory the finite-length flaw than for the infinite-length dynamic fracture toughness model applicabic 1 flaw. Probabilistic analysis of selected I'rS tran-postelcavage arrest conditions in an RPV.
sients produced a substantial decicase in the con-i ditional probability of failure for a finite-length Keency. Walker, J. and llass,11. R., Oak Ridge Na-flaw model. It is recommended that a testing pro-tional Laboratory, "ORNOZL A Finite-Element gram be carried out utilizing clad cruciform and Mesh Generator for Nonic-Cylinder Intersections clad cylindrical test specimens. "Ihc completed Containing inner-Corner Cracks," NUREG/
experimental and analytical research will provide CR-5872, U.S. Nuclear Regulatory Commission, a basis for introducing a refined treatment of sur.
Washington D.C., September 1992.
face flaw initial geometry into l'rS fracture analy-sis procedures. This report is designated liSST This report describes the ORNOZL finite-Report 129.
clement mesh generator program for computa-tional fracture mechanics analysis. The program Kirk, M.T and Dodds, R. I1., University of Illinois,"J automatically generates a three-dimensional and CFOD Estimation Equations for Shallow Cracks (3-D) finibelement model for four different ge-in Single Edge Notch llend Specimens," NUREG/
ometries of a corner crack in a nonle-cylinder CR-5969, U.S. Nuclear Regulatory Commission, I
intersection. ORNOZL generates a core of spe-Washington, D.C., July 1993.
j cial wedge or collapsed prism elements at the Fracture toughness values determined using shal-crack front to mtroduce the appropriate stress low cracked single edge notch bend, SE(II), speci-l l
singularity at the crack tip. Regular 20-noded iso-mens of structural thickness are useful for parametric brick elements are used away from the struct u ral in t egrity assessm ents. Results from two crack front m the modeling. Also, an option is dimensional plane strain finite-element analyses mcluded that allows for an embedded or pene-are used 1o develop J and CTOD estimation strat-trating crack m clad mat erials. As few as five input egies appropriate for application to both shallow cards are required to execute the program. OR-and deep crack SE(II) specimens. Crack depth to NOZL is part of a three-program system,OR-specimen width (a/W) ratios between 0.05 and NOZL-ADIN A-ORVIRT, which addresses hnear 0.70 are modelled using Ramberg-Osgood strain or nonlinear fracture m 2-or 3-D crack geome-hardening exponents (n) between 4 and 50. The tries. ORNOZL creates files contaming nodal estimation formulas divide J and CrOD into pomt coordinates and clement connectivities that small scale yielding (SSY) and large scale yielding have formats compatible with the ADINA struc-(LSY) components. For each case, the SSY com-tural analysis program. ORVIRT ts a post-proces-ponent is determined by the linear clastic stress sor to ADINA and employs a virtual crack intensity factor, Kg.'Ihe formulas differ in evalua-extension technique to compute energy release tion of the LSY component.The techniques con-rates at specified positions along the crack front.
sidered include: estimating J or CTOD from plastic work based on load line displacement (A /
pi Keency-Walker, J., et al., Oak Ridge National Labo-LLD), from plastic work based on crack mouth ratory," Finite-Length Surface Crack Propagation in opening displacement (A /CMOD), and from pi Clad Cylinders," NUREG/CR-5915, U.S. Nuclear the plastic component of crack mouth opening 9
Compilation of Reports-1991-1993 displacement (CMODg). Ag/CMOD provides tile tearing data thus allowing both failure modes the most accurate J cstimation possible.
to be incorporated in a single toughness locus.
l The evolution of Q, as plasticity progresses from Kirk, M. T. and Dodds, R. H., University of I!!inois, small scale yiciding to fully yielded conditions,
" Approximate Techniques for Predicting Size Effects has been quantified for several crack geometries on Cleavage Fracture 1bughness (Jc)," NUREG/
and for a wide range of material strain hardening CR-5970, U.S. Nuclear Regulatory Commission, properties. An indicatorof the robustness of the Washington, D.C., July 1993.
J-Q ficids is introduced; Q as a field parameter and as a pointwise measure of stress level is dis-His investigation examines the ability of an clas-cussed.
l tic T-stress modified boundary layer (Mill) solu-tion to predict stresses ahead of a crack tip in a Rosenfield, A. R. and Marschall, C. W., llattelle Mc-variety of planar geometries. The approximate morial Institute, " Fracture Mechanics Ilased Failure stresses are used as input to estimate the effective Analysis," NUREG/CR-5860, U.S. Nuclear Regula-driving force for cleavage fracture Jo using the tory Commission, Washington, D.C., June 1992.
micromechanically based approach introduced by henty case studies involving the application of Dodds and Anderson. Finite element analyses for fracture mechanics to structural integrity have a wide variety of planar cracked geometnes are been reviewed and compared with a simdar report conducWd which have clastic biaxiahty parame-published in 1978. Sixtecn of the new cases discuss ters ranging from --0.99 (very low constramt) to failures, while four are fitness-for-purpose analy-i
+ 2.% (very high constraint). ne magnitude and ses (i.e., cvaluation of safe operating conditions of sign of indicate the rate at which crack-tip con-defect-containing struct ures). Compared with the straint changes with increasing applied load. All earlier study, no significant improvement in accu-results pertain to a moderately strain hardening racy of failure analysis was detected. Ilowever, material (strain hardening exponent n of 10).
expert opinion suggests that there has been Dese analyses suggest that p is an effective indi-significant improvement in fitness-for-purpose cator of both the accuracy of T-Mill estimates of analysis.
Jnand of applicability limits on evolving fracture analysis methodologies (i.e. T-Mill, J-0, and J/
Shum, D. K., et al., Oak Ridge National Laboratory, Jo). Specifically, when lQl > 0.4 these analyses
" Analytical Studies of 'hansverse Strain Effects on show that t he T-MIllapproximation of Jo is accu-Fracture 7bughness for Circumferentially Oriented rate to within 20% of a detailed finite-element Cracks " NUREG/CR-5592, U.S. Nuclear Regula-analysis. As " structural type" configurations, i.e.
tory Commission, Washington, D.C., April 1991.
l shallow cracks in tension, generally have l@l The objective of this report is to describe the
> 0.4, it appears that only an clastic analysis may development of analysis methods for estimating l
l be needed to determine reasonably accurate Jo the decrease in crack-initiation toughness, from a values for structural conditions.
reference plane strain value, due to positive straining along the crack front of a circumferen.
O'Dowd, N. It, Imperial College, and C. E Shih, Ilrown tial flaw in a reactor pressure vessel. ne analysis University, 'Two Parameter Fracture Mechanics:
methods are based on two different approaches neory and Applications." NUREG/CR-5958, U.S.
that are currently being developed to analyze and Nuclear Regulatory Commission, Washington, D.C.,
explain the effects of transverse strain and stress February 1993.
states on fracture toughness. The first approach relates crack initiation with material failure at A family of self-similar fields provides the two points within a few crack-tip opening displace-parameters required to characterize the full ments directly ahead of the crack tip. In this re-range of high-and low-triaxiality crack tip states.
port the first approach is used to examine ductile ne two parameters, J and Q, have distinct roles:
crack initiation. His micromechanical approach J sets the size scale of the process zone over which thus provides a relation between fracture tough-large stresses and strains develop, while Q scales ness and values of the effective stress and strain at the near.tip stress distribution relative to a high failure that can be determined experimentally.
triaxiality reference stress state. An immediate The second approach focuses on the develop-consequence of the theory is this:it is the tough-ment of correlation parameters that relate frac-ness values over a range of crack tip constraint ture toughness with nominal stress and strain that fully characterize the material's fracture re-states. Candidat e correlation parameters include, sistance. It is shown that Q provides a common but are not limited to, the area enclosed within a scale for interpreting cleavage fracture and duc-selected maximum principal stress contour and NUREG-1426 10
Compilation of Reports-1991-1993 the plastie zone si7c. In the first phase of this ding region; (2) linear-clastle vs elastic-plastic de-work, the scope of the present investigation is scription of material response; and (3) limited to crack-front constraint conditions that base-material-only vs bimaterial cladding-base can be described in terms of conven tional one-pa-vessel-model assumption. The sensitivity evalua-rameter (K orJ),in-planc, near-tip ficids and the tion indicates that the analysis results are very transverse strain.'Ib date, correlation parameter sensitive to the above assumptions. This report is investigations have concer,t ated on the stress designated 11SST Report No.139.
contour method, which is used to examine cicav-age crack initiation. Validation checks of thc anal-ysis methods under study are being made by Fracture Mechanics: Experimental -
applying them to small-and larre-scale fracture Component Testing i
data. Preliminary estimates on the change in crack mitiation toughness associated with either Irwin, G. R., University of Maryland,"Use of Thick-negative or positive straining along a crack front ness Reduction to Estimate Values of K." NUREG/
have been obtamed for conditions of contamed CR-5697, U.S. Nuclear Regulatory Commission, crack-tip yiciding. Results from the validation Washington, D.C., November 1991.
I checks are promising but incomplete. The prima-Using results for two 152-mm-thick wide-plate ry problem encountered in the validation checks tests at the National Institute of Standards and is the absence of accurate descriptions of the Technology, estimates of K were made using the
{
near tip stress and strain ficids associated with residual thickness reduction near the planc of fracture. These results corresponded well to the the generation of some of these fracture data. In addition, there are reasons to believe that two-pa-average of K values for c!cavage arrest and rein-rameter in-plane approaches may be needed to itiation obtained at Oak Ridge National Labora-characterize crack initiation in some of these tory using generation-mode, dynamic-analysis tests. Included are recommendations for subse-computations.
quent phases of the work considered necessary to provide more precise estimates on the effects of Iskander, S. K., et al., Oak Ridge National Laborato-positive out-of-plane straining on the crack-ry," Experimental Results of'Iests to Investigate Flaw initiation toughness for circumferentially ori-13chavior of Mechanically Loaded Stainless Steci 1
ented flaws.
Clad Plates," NUREG/CR-5785, U.S. Nuclear Reg-ulatory Commission, Washington, D.C., April 1992.
i Shum, D. K., et al., Oak Ridge National Laboratory,
{
" Potential Change in Flaw Geometry of a.: Ini.
A small cmck near the inner surface of clad nuclear tially Shallow Finite-Length Surface Flaw During a reactor pressure vessels is an important consider-Pressurizedllhennal-Shock Transient," NUREG/
ation in the safety assessment of the structural in-CR-5(X>S. U.S. Nuclear Regulatory Commission. Wash, tegrity of the vessel. Four-point bend tests on large ington, D.C., September 1993.
plate specimens, six clad and two unciad, were per-formed to determine the effect of stainless steel This study presents preliminary estimates on cladding upon the propagation of small surface whether an initially shallow, axially oriented, in-racks subjected to stress states similar to those pro-ner-surface finite-length flaw in PWR-RPV duced by pressutized thermal shock conditions. Re-would tend to clongate in the axial direction and/
sults of tests at tempcotures 10 and 60'C below the 3
or deepen into the wall of the vessel during a nil-doctility transition temperaturc have shown that postulated I'rS transient. Analysis results ob-a tough surface layer composed of cladding and tained based on the assumptions of (1) linear-heat-affected zone has arrested running flaws in clastic material response, and (2) cladding with clad plates under conditions where unciad plates same toughness as the base metal, indicate that a have ruptured. Furthermore, the h ad-bcaring m-nearly semicircular flaw would likely propagate in pacity of clad plates with large subclad flaws signifi-the axial direction followed by propagation into (nntly exceeded that of an unclad plate with a much the wall of the vessel. Note that these results smaller flaw. More testing is necessary to unambig-correspond to initiation within the lower-shelf uously single out whether it is the cladding or the j
fracture toughness temperature range, and that heat-affected zone that is primanly responsibic for their general vahdity within the lower-transition the observed enhanced kud-bearing (npacity of temperature range remains to be determined.
plates. 'Ihe compressive stresses that limited the The sensitivity of the numerical results and con-depth to which the flaw could propagate are absent clusions to the following analysis assumptions are in a repressurization event. Nonetheless, the ex-cvaluated: (1) reference flaw geometry along the periments show that if the surface layer is sufficient-entire crack front and especially within the clad-ly tough, it could prevent a flaw, near the surface, 11 NUREG-1426
l t
Compilation of Itcports-1991-1993 from propagating along the surfacc. 'lhe flaw could fcct the integrity of RPV supports; and (2) an tunnel below the surface, but a sufficiently tough overall assessment of low upper-shelf (LUS) surface layer would reduce the maximum stress welds in IIPVs with special emphasis on reevalu-intensity factor.
ating ductile tearing criteria. 'lhe first four stub-panel crack-arrest tests wcre performed. Post test Naus, D. J., et al., Oak Ridge National Laboratory, material characterization was performed for clad-
" Crack-Arrest Ilchavior in SEN Wide Plates of Low-plate and wide-plate Series 2 test materials. Sta-Upper-Shelf liase hietal Tested Under Nonisother-tistical analyses were performed on the data from mal Conditions: WP-2 Series," NUREG/CR-5451, the Fifth IISST Irradiation Series on the study of U.S. Nuclear Regulatory Commission, Washington, 14 shifts. Analysis of the irradiated fracture-D.C., August 1990.
toughness testing was completed for the Seventh IISST Irradiation Series on cladding. Detailed Six wide-plate crack. arrest tests (WP-2 Series) planning was begun for the next pressurized-arc discussed in this report. Each test utilized thermal-shock experiment, l'FSE-4, to examine either a 1 x 1 x 0.1-m or a 1 x 1 x 0.15-m thick the extent of ductile tearing and its interaction single-cdge notch specimen (a/W = 0.2), fabri-with c!cavage fracture in an LUS weld metal.
cated from a low-upper-shelf base material, that was subjected to a linear thermal gradient along Pennell, W. E., Oak llidge National Laboratory, the plane of crack propagation. 'lhe tests were "llcavy Section Stect Technology Program. Semian-conducted at the National Institute of Standards nual Progress Report for October 1989 - h1 arch and 'Ibchnology and were designed to pro ide 1990," N U REG /CR-4219, Vol. 7, No.1, U.S. N uclear fracture-toughness measurements at tempera-Regulatory Commission, Washington, D.C., h1 arch tures approaching or above the onset of the Char-1991.
py upper-shelf regime, in a rising toughness
'lh e l icavy Section St ecl'Ibchnology (l lSST) Pro-region, and with an mercasm, g dnymg force. Re-gram is conducted for the Nuclear Regulatory sults obtamed from these tests have produced Commission (N RC)by Gak Ridgc National Labo-crack-arrest toughness values wc!! above the limit ratory (ORNL). The program focus is on the de-recogni/cd by the current ash 1E guidchnes (220 elopment and validation of technology for the h1 Pad ) with arrests occurring at up to 102*C assessment of fracture prevention margins in above the material DW-NDT (60*C). The frac ~
commercial nuclear reactor pressure vessels. In ture data support: (1) use of fracture mechames the current reporting period, reorganii.ation of concepts to analyre cicavage run-arrest events
- the original llSST program into separate pro-(2) treatment of cleavage and ductile fracture grams with emphasis on fracture mechanics tech-modes as separate events, and (3) fact that cleav-nology (IISSO and materials irradiation effects age arrest occurs above the ash 1E limit.
(IISSI) has been completed. 'lhe revised llSST program is organized in 10 Tasks. These are Pennell, W. E., Oak Ridge National Laboratory, (1) Program hianagement, (2) Fracture hiethod-
"IIcavy Section Steel Technology Program. Semian-ology and Analysis,(3) hiaterial Characterization nual Progress Report for April - September 19N9,"
'lhsks, (4) Special Technical Assistance, (5) Crack N U REG /CR-4219, Vol. 6, No. 2 U.S. Nuclear Regu-Arrest Technology. (6) Cicavage Crack Initiation, latory Commission, Washington, D.C., September (7) Cladding Evaluations, (8) Pressurized-1990.
Thermal-Shock Technology, (9) Analysis h1cth-ods Validation, (10) Fracture Evaluation 'Ibsts.
The licasy-Section Steel 'Ibchnology (llSST)
.the program tasks have been structured to place Program studies concern all areas of the technol-emphasis on the resolution fracture issues with ogy of materials fabricated into thick-sectinn, near-term heensm, g sigmficimcc, primary-coolant containment systems of light-water-cooled nuclear power reactors. The focus is Pennell, W. E., Oak Ridge National Laboratory, on the behavior and Structural integrity of stect lleavy Section Stect Technology Program. Semian-reactor pressure vessels (RPVs) contammg crack-nual Progress Report for April - September 1990,"
like flaws. Durmg this penod, analytical efforts N UREG /CR-4219, Vol. 7, No. 2, U.S. Nuclear Regu-meluded exammmg the mfluence of high crack-latory Commission, Washington, D.C., September arrest toughness on RPV integrity and an m-3993*
creased emphasis on evaluating large internation-al structural experiments. Two areas of NRC The ficavy-Section Steel 'Ibchnology (IISST) topical support were continued:(1) the evaluation Program is conducted for the Nucicar Regulatory of mechanisms for enhanced low-temperature, Commission (N RC) by Oak Ridge National Labo-low-flux irradiation embrittlement that may af-ratory (ORNL). The program focus is on the NUREG-1426 12
Compilation of Reports-1991-1993 l
development and validation of technology for the Pennell, W. E., Oak Ridge National Laboratory, assessment of fracture-prevention margins in "licavy Section Steel Technology Program. Semian-commercial nuclear reactor pressure vessels. Re-nual Progress Report for April - September 1991,"
organization of the original iISST Program into NUREO/CR-4219,Vol.8 No.2,U.S.NuclearRegu-separate programs with emphasis on fracture-latory Commission, Washington, D.C., April 1992.
mechanics technology (IISST) and matcrials-irra-diation effects (IISSI) was previously completed.
'Ihe fica y-Section Steel lechnology (IISSV Pro-1he revised IISST Program is organized in 10 gmm is conducted for the Nuclear Regulatory tasks: (1) program management (2) fracture Commission by Oak Ridge National IAoratory methodology and analysis, (3) mat crial character-(ORNL). 'the program focus is on the development ization and properties, (4) special technical assis-and validation of technology for the assessment of tance, (5) crack-arrest technology, (6) cleavage-fracture-prevention margins in commercial nuc! car crack initiation. (7) cladding evaluations. (S) pres-reactor pressure vessels. The liSST Progmm is or-suri/cd-thermal-shock technology, (9) analysis pani /cd m 10 tasks: (1) program management, methods validation, and (10) fracture evaluation (2) foeture methodology and analysis, (3) material tests.'lhe program tasks have been structured to chameterization and properties, (4) special techni-place emphasis on the resolution fracture issues cd assistance, (5) fracture analysis computer pro-with near-term licensing significance. Resources grams. (6) cleavage-emck initiation, (7) cladding to execute the research tasks are drawn from evaluations, (8) pressurized-thermal-shock technol-ORNL with subcontract support from universi-ogy (9) analysis methods validation, and (10) frac-tics and other research laboratories. Close con.
ture evaluation tests lhe program tasks have been tact is maintained with related research programs structured to place emphasis on the resolution of l
both in the United States and abroad, fmeture issues with near-term licensing signifi-cance. Resources to execute the research tasks are Pennell, W. E., Oak Ridge National Laboratory, drawn from ORNL with subcontract support from
" Heavy Section Stect Technology Program. Semian-universitics and other rescan:h labomtories. Close nual Progress Report for October 1990 - March contact is maintained with related research pro-1991," N UREG /CR-4219, Vol. 8, No.1, U.S. Nuclear gmms both in the United States and abroad.
Regulatory Commission Washington, D.C., Febru-ary 1992.
Pennell, W. E., Oak Ridge National Laboratory, "licasy Section Steel 1bchnology Program. Semian-
'Ihc 11 easy-Section Steel Technology (llSST) nual Progress Report for October 1991 - March Program is conducted for the Nuclear Regulatory 1992." N UR EG /CR 4219. Vol. 9, No.1. U.S. Nuclear Commission (NRC)by Oak Ridge National Labo-Regulatory Commission Washington, D.C., Novem-ratory (ORNL). The program focus is on the de-ber 1992.
velopment and validation of technology for the assessrrent of fracture-prevention margins in The 11 easy Section Steel Technology (IISST) commt rcial nuclear reactor pressure vessels. Re-Program is conducted for the Nuclear Regulatory organization of the original iISST Program into Commission (NRC)by Oak Ridge National Labo-separate programs with emphasis on fracture-ratory (ORNL). The program focus is on the de-mechanics technology (HSST) and materials-velopment and validation of technology for the irradiation effects (liSSI) was previously com-assessment of fracture-prevention margins in pleted. The revised liSST Program is organized commercial nuclear reactor pressure vessels.The in 10 tasks: (1) program managem ent, (2) fracture IISST Program is organized in 11 tasks: (1) pro-methodologyandanalysis (3)materialcharacter-gram management; (2) fracture methodology and ization and properties. (4) special technical assis-analysis; (3) mat erial characterization and proper-tance, (5) crack arrest technology, (6) cleavage-tics; (4) special technical assistance; (5) fracture crack initiation, (7) cladding evaluations, (8) pres-analysis computer programs; (6) c!cavage-crack surized-thermal-shock technology, (9) analysis initiation; (7) cladding evaluations; (8) pressur-methods validation, and (10) fracture evaluation ized-thermal-shock technology; (9) analysis tests.1he program tasks have been structured to methods validations; (10) fracture evaluation place emphasis on the resolution fracture issues tests; and (11) warm prestressing. The program with near-term licensing significance Resources tasks have been structured to place emphasis on to execute the research tasks are drawn from the resolution fracture issues with near-term li-ORNL with subcontract support from universi-censing significance. Resources to execute the tics and other research laboratories. Close con-research tasks are drawn from ORNL with sub-tact is maintained with related research programs contract support from universities and other re-f both in the United States and abroad.
search laboratories. Close contact is maintained 13 NUREG-1426
Compilation of Report's-1991-1993 with the sister Heavy-Section Steel Irradiation tions to correlate best with the omega criterion (HSSI) Program at ORNL and with related re-which defines limits on both the maximum J level search programs both in the United States and and the maximum crack extension allowable for a abroad. This report provides an overview of prin-particular specimen size and material toughness cipal developments in cach of the 11 program combination. The final section looks at the prob-tasks from October 1,1991 to March 31,1992.
Icm of extrapolation ofJ-R curve data when necd-ed for a structure fracture analysis. Several forms Fracture Mechanics: Experimental -
f extrapolation relationships are compared from the point of view of accurate and conservative Standard Specimen Testing extrapolation, particularly from the standpoint of Dally, J. W., University of Maryland, et al., " Lower tearing instability analysis of a growing, ductile Bound Initiation 1bughness with a Modified Charpy crack on the material upper shelf.
Specimen," NUREG/CR-5703, U.S. Nuclear Regu-Joyce,J. A., U.S. Naval Academy,et al.," Comparison latory Commission, Washington, D.C., November ofJei and J-R Curves for Short Crack and Tensilcly 199L Loaded Specimen Geometrics of a High Strength
" Lower-bound" initiation toughness of A 533 B Structural Steel," NUREG/CR-5879, U.S. Nuclear reactor grade steel was determined over the tem-g[ tory Commission, Washington, D.C., Novem-g perature range from 0 to 57'C by using a
~~
modified Charpy specimen. The lower-bound This paper describes an experimental program measurements were attained by utilizing the fol-which had the objective of developing a series of lowing procedures: (1) dynamic loading, (2) modi-J R curve data from laboratory specimens of var-fication of the geometry of the specimen, and ied constraint. Constraint was varied by testing (3) axial precompression of the notch.1he report specimens with different thicknesses, crack describes in detail the key features of the modi-lengths, and mode ofloading. All specimenswere ficd geometry, the method of precompressing the relatively small and were kept simple in geometry specimens, and the strain-gage procedure. The and loading to allow estimation of the applied J dynamic initiation toughness K, which corre-integral. All tests were conducted on high id lates with the lower-bound toughness, was deter-strength structural steel, at ambient temperature, mined by analyzing strain-time records from the on the ductile upper shelf for this alloy. Results of specimen.'Ihe results from a fractogaphic analy-these tests have shown that different constraint sis were correlated with those from the strain.
condition can dramatically affect the J c, and the i
time analysis. An empirical correlation was J-R curve for the full range of crack lengths and developed relating K to the energy absorbed Eev loading modes studied here. The results are com-i during the fracture of the specimen. Finally, the pared in terms of the "T Stress" Tg parameter lower-bound toughness from this study compared and the O constraint parameter, but the trends in favorably with K and Kid measurements from the the data do not seem to correlate well with cither i
same material established in other programs.
parameter. Although both theT and Q parame-y ters predict that the single edge notched tensile Joyce, J. A., U.S. Naval Academy, and Hackett, E.
har SE(T) would have relatively high constraint, M., David laylor Research Center " Extension and this geometry demonstrated the highest J e prop-i Extrapolation of J-R Curves andTheir Application to crties. The double edge notched bars were pre-the Low Upper Shelf "loughness Issue," NUREG/
dicted by the T or Q parameters to be a low CR-5577, U.S. Nuclear Regulatory Commission, y
Washmgton, D.C., March 1991.
constraint geometry, however this geometry re-suited in J ei results lower than those measured on This document develops methods of measuring standard deeply crack bend bars.
cxperimentally the limits of valid fracture me-chanics data that can be obtamed from small frac-Joyce, J. A., U.S. Naval Academy, and Link, R. E.,
Naval Surface Warfarc Center "The Effect of Elec-ture mechanics specimens. Ihc proposed technique generally shows that present ASTM tric Discharge Machined Notches on the Fracture limits arc overly conservative and the new tech-lbughness of Several Structural Alk>ys," NUREG/
nique would allow almost a three-fold increase in CR-5981, U.S. Nuclear Regulatory Commission, the amount of crack extension allowed in the test-Washington, D.C., September 1993.
i ing of a surveillance specimen. Analytic relation-Recent computational studies of the stress and ships are then developed to allow use of the new strain ficids at the tip of very sharp notches have experimentally measured limit to J controlled shown that the stress and strain fields are very crack growth for design or failure analysis apphca-weakly de;3cndent on the initial geometry of the NUREG-1426 14
Compilation of Reports-1991-1993
(
i notch once the notch has been blunted to a radius McCabe, D. E., Oak Ridge National Laboratory, that is 6 to 10 times the initial root radius. It "A Comparison of Weibull and pic Analysis of Wansi-follows that if.the fracture toughness of a material tion Range Fracture Tbughness Data," NUREG/
is sufficiently high so that fracture initiation does CR-5788, U.S. Nuclear Regulatory Commission, not occur in a specimen until the crack-tip open.
Washington, D.C., January 1992.
ing displacement (CTOD) reaches a value from 6 to 10 times the size of the initial notch tip diame-Characteristics of extremal statistics that are used ter, then the fracture toughness will be indepen-to predict size effects on cleavage fracture tough-dent of whether a fatigue crack or a machined ness in the transition range were explored. A 533 notch served as the initial crack. In this exper.
grade B steel base and weld metals were tested imental program the fracture toughness (JreandJ using compact specimens ranging in size from resistance (J-R) curve, and CTOD) for several 1/ZrC(T) to STC(T) and with sufficient replica-structural alloys was measured using specimens tion in some cases to provide goed fits to Weibull with conventional fatigue cracks and with EDM distributions. He classical speci,nen size effect machined notches. He results of this program on data scatter and median Kje toughness at a have shown, in fact, that most structural materials gyn tut temperatum wassseMn melowdo n
ns, n rang e s wem wcH do not achieve initiation CLOD values on the predicted with extremal statistics. Ilowever, the order of 6 to 10 times the radius of even the same model is not applicable on the lower shelf smallest EDM notch tip presently achievable. It is and it also becomes extremely weak and unreli-found furthermore that tougher materials do not able in the mid-to high-transition rarge. De seem to be less dependeM on the type of notch tip Irwin pc - pre relationship was also explored as a l
present. Some materials are shown to be much model and was found to predict similar t.ize ef-more dependent on the type of initial notch tip fccts. De predictive characteristics of the latter used, but no simple pattern is found that relates seemed better suited to deal with the diminution this observed dependence to the material of size effectsin the near tolow-shelf toughness strength, tou$ ness, or strain hardening rate.
range. In the rising toughness part of the transi-tion, the predictive cha acteristics were about the same as the statistical model up to where sc (Sje in this study) of the baseline (small specimen) data Landes, J. D., University of 1bnnessee, "Extrapola-were rr or less. His work could be used in the tion of the J-R Curve for Predicting Reactor Vessel establishment of a framework for transition tem.
Integrity," NUREG/CR-5650, U.S. Nuclear Regula-perature test criteria. Upper-and lower-bound tory Commission, Washington, D.C., January 1992.
(Eje criteria could be used to define optimum conditions for the application of either of the aforementioned models. For surveillance pro-i ne work in this report was conducted in support grams, sensible rules should be specified as to of the issues studied by the U.S. Nuclear Regula.
specimen size requirements and numbers of spec-tory Commission Jgy/Ju Working Group during imens to be tested in order to apply these analyti-the period 1987-1989. The major issues studied cal models. Another need would be the definition were the J-R curve extrapolation techniques for of a procedure for the Weibull distribution fitting.
using small-specimen test results to predict duc-
'Ihe present repon suggests items to be consid-e r mpamen@ appkaton oMuc pm.
tile instability in larger structures where the ex-tent of crack extension from the small-specimen test was not sufficient. An additional issue was raised during the course of Ihis work by the testing McCabe, D. E., Oak Ridge National Laboratory, of a low-upper-shelf A 302 steel.The results from
" Evaluation of Crack Pop-ins and the Determination these tests were not typical of ductile fracture in of Their Relevance to Design Considerations,"
many steels and suggested that small-specimen NUREG/CR-5952, U.S. Nuclear Regulatory Com-J R curves may not predict the behavior of large mission, Washington, D.C., February 1993.
structures in some cases.The cause of this behav-ior was studied as wc!! as the consequences of De issue with regard to crack pop-ins is to deter-using the J-R curve results from small specimens mine if such events are significant to design con-of this kind of material. Finally, a discussion and siderations. He literature contains ample recommendations are given relating to the use of evidence of pop-in occurrences, but scant infor-extrapolated J-R curves, mation is offered on how pop-ins should be han-15 NUREG-1426
Compilation of Reports-1991-1993 t
died as an issue for design problems. Ilecause NUREG/CR-5886, U.S. Nuclear Regulatory Com-there are two types of cleavage crack origins, the rnission, Washington, D.C., July 1992.
f problem was subdivided into two classes of mate-rials, monolithic and weldments with brittle zones.The weldment situation can be analyzed as
'lhe Ileavy-Section Stcel Technology (HSST) a crack-arrest toughness capability problem, fol.
Program is investigating the increase in effective lowing the recommendations of Sumpter et al.
fracture toughness of A 533 Il steel associated For monolithic materials, pop-ins are more dan.
with shallow flaws and the implications of the gerous, since they appear to be a part of the more shallow-flaw effect on reactor pressure vessel commonly encountered full-cleavage Kje insta.
(RPV) life assessments. 'Ibst data from beams l
bility distributions. A recommendation is made indicate a significant increase in the fracture l
on how to determine if pop-in events lie ou tside of toughness of shallow-crack specimens compared thc larger body of K;c instabilities.The evaluation with deep-crack specimens in the transition re-procedure recommended by the American Soci.
gion of the toughness curve for unirradiated A cty for Testing and Materials for pop-ins seems to 53311 stect. If the toughness increase present in dismiss the possibility that small crack jumps can the test specimens were also present m a reactor be a safety related issue.The present work sug.
vessel, the impact on pressurized-thermal shock j
gests that nearly all pop-in events, regardless of (PFS) analyses could be significant. 'Ib facilitate the magnitude of crack jump, are relevant to safe.
transferability of the specimen data to an RPV, l
ty issues.
post-test finite-clement analyses have been per-formed on stveral test specimens and a reactor vessel for a single (l'rS) transient. 'Ihe analyses j
Rolfe, S. T., University of h,ansas, "The Behavior of are sufficiently refined to allow interpretation of Shallow Flaws m Reactor Pressure Vessels,
the resulte in terms of the J-integral and the so-
[
l NUREG/CR-5767, U.S. Nuclear Regulatory Com-called Q-stess parameter under planc-strain mission, Washington, D.C., November 1991, analysis assuraptions. A negative Q-stress param-eter is indicat.ve of a loss of crack-tip constraint.
.The objective of th.is report is to recommend which is associated with an increase in the frac-those rescarch mvestigations that are necessary ture toughness. Analyses of the test specimens i
to understand the phenomenon of shallow behav-indicate that at the onset of crack initiation the i
nor as it affects fracture toughness so that the deep-crack specimens exhibit an essentially zero results etm be used properly m the structural mar-Q-stress parameter but that the shallow-crack
[
gm assessment of reactor pressure vessels (RPVs) specimen exhibits a Q-stress parameter of about r
with flaws. Preliminary test results of A 533 II
-0.7, which indicates a substantial loss of con-l steci show an elevated crack-tip opening displace-straint in the shallow-crack beam. Using the test ment toughness similar to that observed for struc-data and post-test analysis, a k>cus of toughness tural steels tested at the Um,versity of Kansas.
data in terms of the J. integral and the Q-stress i
Ihus, the mherent resistance to fracture imtia*
parameter has been constructed for a particular tion of A 53311 stcel with shallow flaws appears to temperature. Analyses were also performed on be higher than that used in the current Amencan an RPV with a shallow flaw under FFS loading Society of Mcchamcal Engineers design curves conditions up to the maximum value of J. At maxi-based on testing fracture-mechanics specimens mum J, the analyses reveal a Q-stress parameter with deep flaws. If this higher toughness of labo-about -0.2 to -0.4, which indicates some con-ratory specimens with shallow fiaws can be trans-straint loss but less than in the shallow-rack test ferred to a higher resistance to failure m RPV specimens. Considering the RPV in terms of J-in-design or analysis. Ihen the actual margin of safe-tegral and Q-stress suggests there may be a larger ty in nuclear vessels with shallow flaws would be margin of safety than would be found using the I
greater than is currently assumed on the basis of J-integral alone. Thermal-shock data, which were deep-flaw test results. 'lhis report reviews those generated using cylindrical vessels under thermal factors and makes recommendauons of studies shock loading, show no significant increase in that are needed to assess the transferability of toughness even for shallow-flaw depths. 'The i
shallow-flaw toughness test results to the struc-thermal shock data seem to indicate two offset-tural margm assessment of RPV with shallow ting effects: a shallow-flaw effect, which increases i
flaws.
toughness, and an out-of-plane (biaxial) stress ef-fect, which decreases toughness. Additional work
'lheiss,'E J., Oak Ridge National Laboratory, et al.,
is necessary to resolve outstanding issues for ap-
" Experimental and AnalyticM Investigation of the plying shallow-crack data to an RPV and validat-l Shallow - Flaw Effect in Reactor Pressure Vessels,**
ing the J-Q technique for fracture evaluations.
j NUREG-1426 16
Compilation of Reports-1991-1993 Theiss,1: J., et al., Oak Ridge National Laboratory, Northwest Lalmratory, operated for the Depart-
" Initial Results of thc Influence of Hiaxial Loading on ment of Energy by Hattelle Memorial Institute.
Fracture 'Ioughness," NUREG/CR-6036, U.S. Nu-The program has shown the feasibility of continu-clear Regulatory Commission, Washington, D.C.,
ous, on-line AE monitoring to detect crack June 1993.
growth and produced validated methods for ap-A testing program to examine the influence of plying the technology. Included are relationships for estimating flaw severity from AE data and biaxialloads on the fracture toughness of shallow-field applications at Watts Har Unit 1 Reactor, flaw specimens under conditions prototypic of a Limerick Unit 1 Reactor, and the High Flux Iso-reactor pressure vessel was begun. Existing data tope Reactor. This report discusses the program suggest that shallow-flaw specimens underbiaxial loading will exhibit a toughness reduction com-scope and organization, the three program phases and the results obtained, standard and code acti-pared to comparab!c uniaxial specimens. Quan-vities, and instrumentation and software devel-tification of this toughness reduction is the main oped under this program.
goal of the biaxial fracture toughness program. A cruciform specimen with a two-dimensional shal-low through. thickness flaw under a biaxial load NDE - In-Service InsI>ection ratio of 0.6:1 was used for biaxial fracture tough-Doctor, S. R., et al., Pacific Northwest Laboratory, ness testing. 'lhe critical fracture load for each
" Nondestructive Examination (NDE) Reliability for specimen was approximately the same, but the Insenice Inspection of Light Water Reactors. Semi-uniaxial specimen withstood substantially more annual Report for April - September 1989,"
deformation at failure than did the biaxial speci-NUREG/CR-4469, Vol.11 U.S. Nuc! car Regulatory mens. 'Ihree-dimensional, clastic-plastic, fi-Commission, Washington, D.C., August 1991.
nite-element post-test analyses were necessary to estimate fracture toughness. In all cases, agree.
Evaluation and Improvement of NDE Reliability ment between the measured and computed load for Insenice Inspection of Light Water Reactors vs deformation responses was excellent.1bugh.
(NDE Reliability) Program at the Ibcific Northwest ness values for the cruciform specimens were Laboratory was estabbshed by the Nuclear Regula-compared with data from previously tested, deep.
tory Commission to determine the reliability of cur-and shallow-crack specimens. Results from these ren' insenice inspection (ISI) techniques and to tests indicate that the shallow-crack toughness develop recommendations that will ensure a suit-increase is partially, but not totally, removed by ably high inspection reliability.1he objectives of this the application of biaxial loading. liowever, addi.
progmm include determining the reliability of ISI tional data are required to solidify these conclu.
performed on the primary systems of commercial sions. A proposed test matrix for additional light-water reactors (LWRs): using probabilistic uniaxial and biaxial testing is described. This re.
fracture mechanics analysis to determine the im-port has been designated liSST Report No.138.
pact of NDE unreliability on system safety; and evaluating reliability improvements that can be NDE - Continuous Monitorinb, chieved with improved and admnced technology.
A final objective is to formulate recommended resi-(Acoustic Emission) sions to ASME Code and Regulatory requirements, based on material properties, senice conditions, ilutton, P. IL, et al., Ibcific Northwest Laboratory, and NDE uncertainties.1he program scope is lim-
" Acoustic Emission / Flaw Relationships for Inservice ted to ISI of the primaiy systems including the Monitoring of LWRS," NUREG/CR-5645 U.S. Nu-piping, vessci, and other components inspected in c! car Regulatory Commission, Washington, D.C.,
accxirdance with Section XI of the ASME Code.
October 1991.
This is a progress report covering the programmatic The program concerning Acoustic Emission / Flaw work from April 1989 through September 1989.
Relationships for Insenice Monitoring of LWRs was initiated in FY76 with the objective of validat-Doctor, S. R., et al., Pacific Northwest Laboratory, ing the application of acoustic emission (AE) t Nondestructive Exarnination (NDE) Reliability for monitor nuclear reactor pressure-contatmng insenice Inspection of Light Water Reactors. Semi-components during operation to detect crack.ing.
annual Report for October 1989 - September 1990,"
The program has becri supported by the U.S. Nn-NUREG/CR-4469,Vol.12 U.S.NuclearRegulatory cicar Regulato:y Cc mmission, Ofhce of Nuclear Commission, Washington, D.C., May 1992.
Regulatory Research with supplemental support The Evaluation and Improvement of NDE Reli-from the 1bnnessee Valley Authority. Research ability for Inservice Inspection of Light Water and development has been performed by Ibcific Reactors (N D E Reliability) Program at the Pacific l
l l
17 NUREG-1426
l
[
Compilation of Reports-1991-1993 l
Northwest Laboratory was established by the Nu-Doctor, S. R., et al., lheific Northwest Laboratory, I
c! car Regulatory Commission to determine the
" Nondestructive Examination (NDE) Reliability for reliability of current inservice inspection (ISI)
Inservice Inspection of Light Water Reactors. Semi-i techniques and to develop recommendations that annual Report for April 1991 - September 1991,"
l will ensure a suitably high inspection reliability.
NUREG /CR-4469, Vol.14, U.S. N uclear Regula tory
'the objectives of this program include: determin-Commission, Washington, D.C., July 1992.
ing the reliability of ISI performed on the primary
.Ihe Evaluation and improvement of NDE Reli-systems of commercial light-water reactors ability for Inservice Inspection of Light Water l
(LWRs); using probabilistic fracture mecharucs Reactors (NDE Reliability) Program at the lbeific analysis to determine the impact of NDE unrch-Northwest Laboratory was established by the N u-j ability on system safety; and es aluating reliability clear Regulatory Commission to determine the improvements that c:m be aclueved with im-reliability of current inservice inspection (ISI) proved and advanced technology. A final objec-techniques and to develop recommendations that live is to formulate recommended revisions to will ensure a suitably high inspection reliability.
ASM E Code and Regulatory requirements, based
.N o e ivesof thisprogramincludedetermin-on material properties, senice conditions, and ing the reliability of ISI performed on the primary NDE uncertainties. I he program scope is limited
, gg gg to ISI of Ihe primary systems including the pipmg, (LWRs); using probabilistic fracture mechanics f
vessel, and other components mspected in accor-analysis to determine the imEact of NDE unreli-I dance with Section XI of the ASME Code. 'this is ability on system safety; and evaluating reliability a progress report covering the programmatic improvements that can be achieved with im-work from October 1989 through September pr ved and advanced technology. A final objec-1990' tive is to formulate recommended revisions to ASM E Code and Regulatory requirements, based on material properties, senice conditions, and Doctor, S. R., et al., Ibcific Northwest Laboratory, NDE uncertainties.The program scope is timited
" Nondestructive Exammatson (NDE) Reliability for to ISI of the primary systems including the piping.
Insenice Inspection of Light Water Reactors. Semi-vessel, and other components inspected in accor-annual Report for October 1990 - March 1991,"
dar.:e with Section XI of the ASME Code.Thisis NUREG/CR-4469, Vol.13, U.S. Nuclear Regulatory a progress report covering the programmatic Commission, Washington, D.C., July 1992.
work from Apnl 1991 through September 1991.
The Evaluation and improvement of NDE Reli-pctor, S. R.; et al., Ibcific Northwest Laboratory, Nondestructive Exammation (NDE) Reliability for ability for Insenice Inspection of Light Water Insenice Inspection of Light Water Reactors. Semi.
Reactors (N DE Reliability) Program at the Itcific annual Report for October 1991 - March 199.,
Northwest Laboratory was established by the Nu-NUREG/CR-4469, Vol.15, U.S. Nuclear Regulatory clear Regulatory Commission to determme the Commission, Washington, D.C., September 1993.
reliability of current insenice inspection (ISI) techniques and to develop recommendations that The Evaluation and Improvement of NDE Reli-will ensure a suitably high inspection reliability, ability for Insenice Inspection of Light Water The objectives of this program include determin-Reactors (NDE Reliability) Program at the Ibeific ing 1he reliability of ISI performed on the primary Northwest laboratory was established by the Nu-systems of commercial light-water reactors clear Regulatory Commission to determine the (LWRs); using probabilistic fracture mechanics reliability of current inservice inspection (ISI) analysis to determine the impact of NDE unreli-techniques and to develop recommendations that ability on system safety; and evaluating reliability will ensure a suitably high inspection reliability, improvements that can be achieved with im-The objectives of this program include determin-proved and advanced technology. A final objec-ing the reliability of ISI performed on the primary tive is to formulate recommended revisions to systems of commercial light-water reactors ASME code and Regulatory rcquirements, based (LWRs); using probabilistic fracture mechanics on material properties, senice conditions, and analysis to determine the impact of NDE unreli-NDE uncertainties.The program scope is limited ability on system safety; and evaluating reliability to ISI of the primary systems including the piping, improvements that c:m be achieved with im-vessel, and other components inspected in accor-proved and advanced technology. A final objec-dance with Section XI of the ASME Code.This is tive is to formulate recommended revisions to a progress report covering the programmatic ASME Code and Regulatory requirements, based work from October 1990 through March 1991.
on material properties, senice conditions, and NUREG-1426 18
Compilation of Reports-1991-1993 NDE uncertainties.The program scope is limited the primary mode'of fracture with ductile crack to lSl of the primary systems including the piping, extension intervening only during the last few vasel, and other components inspected in accor-cycles of loading.
dance with Section XI of the ASME Code. This is i
a progress report covering the programmatic Kassir, M. K., et al., Brookhaven National Labora-work from October 1991 through March 1992.
tory," Analysis of Crack Initiation and Growth in the liigh Level Vibration Test at Thdotsu," NUREG/
licasler, P. G., et al., Ibcific Northwest Laboratory, CR-6078, U.S. Nuclear Regulatory Commission,
" Statistically Based Reevaluation of PISC-II Round Washington, D.C., August 1993.
Robin Test Data," NUREG/CR-5410, U.S. Nuclear Regulatory Commission, Washington, D.C., May The fligh Level Vibration Test data are used to 1993.
assess the accuracy and usefulness of current en-gineering methodologies for predicting crack ini-1his report presents a re-analysis of an internation.
tiation and growth in a cast stainless steel pipe al PISC-Il round-robin inspection results using cibow under complex, large amplitude hiading.
formal statistical techniques to account for exper.
The data were obtained by testing at room tem-imental error. 'lhe analysis examines: U.S. team perat ure a large scale modified model of one loop performance vs. other participants performance; of a PWR primary coolant system at thelhdotsu flaw sizing performance and errors associated with Engineering Laboratory in Japan. Fatigue crack flaw sizing; factors influencing flaw detection proba.
initiation time is reasonably predicted by applying bi'ity; and performance of all participants with re.
a modified local strain approach (Coffin Mason-spect to recently developed ASME Section XI flaw Goodman equation)in conjunction with Miner's detection performance demonstration require.
rule of cumulative damage. Three fracture me-ments, and develops mnclusions concerning ultm.
chanics methodologies are applied to investigate sonic inspection capability.
the crack growth behavior observed in the hot leg of the model. These are: the AK methodology (Ibris law), aJ concepts and a recently developed Piping limit load stress-range criterion.1he report in-cludes a discussion on the pros and cons of the Joyce, J. A., U.S. Naval Academy, and IIackett, E.
analysis involved in each of the methods, the role M., David 1bylor Research Center, " Elastic - Plastic played by the key parameters influencing the for-Characterization of a Cast Stainless Steel Pipe Elbow mulation and a comparison of the results with the Material," NUREG/CR-5774, U.S. Nuclear Regula-actual crack growth behavior observed in the vi-tory Commission, Washington, D.C., January 1992, bration test program. Some conclusions and recommendations for improvement of the meth-Tests conducted in Japan as part of the High Level odologies are also provided.
Vibration Test (HLVF) program for reactor pip-ing syst ems revealed fatigue crack growth in a cast Paul, D. D., et al., Hattelle, Memorial Institute, stainless steel pipe elbow. The material tested
" Evaluation and Refinement of Leak Rate Estima-was equivalent to ASME SA351-CF8M.1he Da-vid 'Paylor Research Center (DTRC) was tasked to tion Models " NUREG/CR-5128. U.S. Nuclear Reg-1 develop the appropriate material property data to ulatory Commission, Washington, D.C., April 1991.
characterize cyclic deformation, cyclic clas.
Leak-rate estimation models are important cle-tic-plastic crack growth and ductile tearing resis-ments in developing a leak-before-break method-tance in the pipe elbow material. It was found that ology in piping integrity and safety analyses.
the cast stainless st cel was very resistant to ductile Existing thermal hydraulic and crack-opening-h crack extension. J-R curves essentially followed a area models used in current leak-rate estimations blunting behavior to very high J levels. Low cycle have been incorporated into a single computer fatigue crack growth rate data obtained on this code for leak-rate estimation.1he code is crdled material using a cyclic J integral approach was SQUIR1, which stands for Seepage Quantifica-consistent with the high cycle fatigue crack tion of Upsets In Reactor Tbbes. The SQUIRT growth rate and with a standard textbook correla-program hasbeen validated by comparing its ther-tion equation typical for this type of material.
mal hydraulic predictions with the limited exper-Evaluation of crack closure effects was essential imental data that have been published on to accurately determine the crack driving force two-phase flow through slits and cracks, and by for cyclic clastic-plastic crack growth in this mate-comparing its crack-opening-area predictions rial. SEM examination of several of the cyclic J with data from the Degraded Piping Program. In test fracture surfaces indicated that fatigue was addition, leak-rate experiments were conducted 19 NUREG-1426
Compilation of Reports-1991-1993 1
to obtain validation data for a circumferential of seismic loading rates on cracked piping sys-I fatigue crack in a carbon steel pipe girth weld.
tems, Progress for through-wall-cracked pipe in-volved: (1) conducting a 28-inch diameter l
Wilkowski, G. M., et al., llattelle, "Short Cracks in stainless steel SAW and 4-inch diameter French Piping and Piping Welds. Semiannual Report for TP316 experiments, (2) conducting a matrix of l
March - September 1990," NUREG/CR-4599, FEM analyses to determine GE/EPRI functions Vol.1. No.1, U.S. Nuclear Regulatory Commission, for short 'lWC pipe, (3) comparison of uncracked Washington, D.C., May 1991.
pipe maximum moments to various analyses and FEM solutions, and (4) development of a J-esti-
'Ihis is the first semiannual report of the U.S. Nu-mation scheme that includes the strength of both c! car Regulatory Commission's Short Cracks in Pip-the weld and inse metals. Progress for surface-ing and Piping Welds research program. 'lhe cracked pipc involved: (1) conducting two experi-i program began in March 1990 and will extend for 4 ments on 6-inch diameter (Sch. 40 and XXS) pipe l
years. 'Ihc intent of this program is to verify and withd/t - 0.5and elrr = 0.25 cracks,(2)compari-improve fracture analyses for circumferentially sons of the pipe experiments to Net Section-Col-cracked large-diameter nuclear piping with crack lapse predictions, and (3) modification of the sizes typically used in leak-before-break analyses or SC.TNP and SC.TKP J-estimation schemes to in scryice flaw evaluations. Only quasi-static load-include external surface cracks. high-tempera-ing rates are evaluated since the NRC's Interna-ture hardness testing appears to be a useful tional Piping Integnty Research Group (IPIRG) screening eriteria parameter for assessing the sus-progmm is evaluating the effects of seismic kiading ceptibility of ferritic pipe to dynamic strain aging.
rates on cracked piping systems.
For anisotropic fracture evaluations, it was found i
that only one of five ferritic pipes had the low Additional efforts involve investigating phenom-toughness direction in a helical direction, the rest I
ena discovered during the course of conducting had low toughness in the axial direction. For the Degraded Piping program.1hese include the crack-opening area analyses, predictive capabili-evaluation of the occurrence of unstable crack ties were expanded so that load versus crack jumps in ferritic steels at LWR temperat ur es, and opening can be calculated from the LBII.NRC, I
the occurrence of anisotropic fracture propenies GE/EPRI, LBII.GE, LilB.ENG, and Thda/ Paris i
causing helical crack growth. Both of these phe-analyses.1hese include loading due to tension, nomena may affect the safety margins implicit in bending, and combined tension and bending.The LBil analyses. Other investigations deal with the LHB.ENG analysis was also modified to account fracture behavior of bi-metallic welds, and im-for the weld and base metals strengths Elastic provements in crack opening area analyses used FEA showed that for pressure loading, a crack in LHB. Since much of the work in this program close to a terminal end (i.e., a nozzle) will have was just beginning during this first reporting peri-lower crack opening due to restraint of the in-i od and progress is limited, a complete statement duced bending.This could affect LBB analyses.
of work for the whole program is provided in this "E
Wilkowski, G. M., et al., Battelle, "Shon Cracks in I
Piping and Piping Welds. Semiannual Report for Wilkowski, G. M., et al., Battelle, "Short Cracks in April. September 1991," NUREG/CR-4599, Vol. 2, t
Piping and Piping Welds. Semiannual Report for Oc-No.1. U.S. Nuc! car Regulatory Commission, Wash-i tober 1990 - March 1991," N UREG/CR-4599, Vol.1, ington, D.C., September 1992.
No. 2, U.S. Nuclear Regulatory Commission Wash-ington, D.C., April 1992.
This is the third semiannual report of the U.S.
3 Nuclear Regulatory Commission's Short Cracks 1his is the sec(md semiannual report of the U.S.
in Piping and Piping Welds research program.
Nuclear Regulatory Commission's Short Cracks This 4-year program began in March 1990. 'lhe in Piping and Piping Welds researcli program.
overall objective of thts program is to verify and The program began in March 1990 and will extend improve fracture analyses for circumferentially for 4 years.'Ihe intent of this program is to verify cracked large-diameter nuclear piping with crack and improve fracture analyses for circumferen-sizes typically used in leak-hefore-break analyses tially cracked large-diameter nuc! car piping with or inser ice flaw evaluations.
crack sizes typically used in leak-before-break analyses or in. service flaw evaluations. Only qua-si-static h)ading rates are evaluated since the Wilkowski, G. M., et al., Battelle, "Short Cracks in
)
NRC's International Piping Integrity Research Piping and Piping Welds. Semiannual Report for Oc-Group (IPIRG) program is evaluating the effects tober 1991 - March 1992." NUREG/CR-4599, Vol. 2, NUREG-1426 20
Compilation of Reports-1991-1993 No. 2, U.S. Nuc! car Regulatory Commission, Wash-would indicate fracture behavior different from ington, D.C., May 1992.
that observed in bulk-r laterial fracture tests.The A 508 material in the liAZ region, very close to This is the fourth semiannual report of the U.S.
th e welds, contains smtll (3 mm) regions adjacent Nuclear Regulatory Commission's Short Cracks to each layer of weld runs where grain coarsening in Piping and Piping Welds research program.
and hardness elevaiion suggest reduction of This 4-year program began in March 1990. The cleavage initiation tc ughness. The degree of se-overall objective of this program is to verify and verity is largest wheie this hical region coincides improve fracture analyses for circumferentially with a local elevation of carbide density in the A cracked large-diameter nuclear piping with cmck 508 material. The A 508 II AZ region adjacent to sizes typically used in leak-before-break analyses the topmost weld run may be the region most or in-service llaw evaluations. Progress during likely to assist cleavage-fracture initiation be-t his reporting period involved: (1) completing two cause of its k) cation: close to a free surface, small through wall-cracked pipe experiments and cracks, and the 11 AZ region beneath the cladding.
I supplementary material property data,(2) an in-It was noted that the small cracks under the clad-ternal circumferential surface-cracked pipe ex-ding have the appearance of prior austenite grain periment was completed which showed that the boundary separations that connect to austenite R/t effects on the Net-Section-Collapse pre-grain boundaries in the cladding. The extreme dicted loads for surface-cracked pipe to be inde-hardness of a narrow layer of cladding at the fu-pendent of crack size, (3) the anisotropy sion boundary may be of interest in further stu-investigation showed that pipe dimensions may be dies of cladding toughness.
as important in determining the out-of-plane crack growth angle as the anisotropyof the tough-ness, (4) we initiated a probabilistic analysis of Nanstad, R. K., et al., Oak Ridge National Laborato-L1311 to assess the potential changes in the leak.
ry, " Chemical Composition and RTNOT Determina-age detection criteria in NRC Reg Guide 1.45, tions for Midland Weld WF-70," N UREG/CR-5914, and (5) other efforts involved a sensitivity study U.S. Nuc! car Regulatory Commission, Washington, on the effect of thermal aging of cast stainless D.C., December 1992.
steel on the moment-carrying capacity of the pipe as a function of time.
'lhe Heavy-Section Steel Irradiation Program 7bnth irradiation Series has the objective to in-vesti ate the cUccts of radiation on the kactum 8
Pressure Vessel Steels toughness of the low-upper-shelf submerged-arc Dally, J. W., University of Maryland, et al., "The welds (ll&W designation WF-70)in the reactor influence of Precompression on the lower-Ilound pressure vessel of the canceled Midland Unit 1 Initiation Tbughness of A 533 Il Reactor Grade
".xlear plant Ws mport discusses determina.
Steel," NUREG/CR-5847, U.S. Nuclear Regulatory tion of variations m chemical composition and Commission, Washington, D.C., May 1992, reference temperature RTNDT throughout the welds. Specimens were machmed from different This report first describes the test method em.
sections and through thickness locations in both phiying a precompressed round bar subjected to th e beltline and nonle course welds. ne nil-due-impact loading to initiate a cleavage fracture.The tility transition temperatures ranged from --40 to procedure to convert strain measurement into
-60'C (-40 and -76'F) while the RT on con-N dynamic initiation toughness Kid is described.
trolled by the Charpy behavior, varied from -20 to Also, the results of a fractographic analysis are 37'C (-4 to 99'F).The upper-shelf energies var-correlated with the features observed on the ied from 7710108 J (57 to 80 ft-lb). He combined strain-time traces, and techniques used to distin.
data revealed a mean 41-J (30-ft-Ib) temperature guish initiation by either cleavage or ductile tear.
of-8'C(17'F)with a mean upper-shelf energy of ing are presented.
88 J (65 ft-Ib). The copper contents range from 0.21 to 0.34 wt % in the beltline weld and from lewin, G. R., and Zhang. X. J., University of Mary.
0.37 to 0.46 wt % in the nonle course weld. Atom land," Gradient Study of a Large Weld Joining Two probe field ion microscope analyses indicated Forged A 508 Shells of the Midland Reactor Vessel,"
substantial depletion of copper in the matrix but NUREG/CR-SS67, U.S. Nuclear Regulatory Com.
no evidence of copper clustering. Statistical anal-l mission, Washington, D.C., June 1992.
yses of the Charpy and chemical composition re-I sults as well as interpretation of the ASME l
The low-carbon welds (WF67 and WF70) in the procedures for KrNDT determination are dis-slab examined contained no abnormalities that cussed.
21 NUREG-1426
Compilation of Reports-1991-1993 Radiation Embritilcment in test vs power reactor environments was ob-served for the weak test orientation (ASTM leC)
Chopra, O. K., Argonne National Laboratory, and whereas correspondence was good for the strong Rosinski, S.T., Sandia National Laboratories,"Radi-orientation (ASTM C-L).'lo resolve the anomaly ation Embrittlement of the Neutron Shield 1hnk directly, Charpy-V specimens from a low (essen-from the Shippingport Reactor," NUREG/CR-5748, tially-nil) fluence region of the vessel were irra-U.S. Nuclear Regulatory Commission, Washington, diated together with archive material at 279'C in D.C., October 1991.
the U13R test reactor. Properties tests before UllR irradiation revealed a significant difference
'the irradiation embrittlement of Shippingport in 41-J transition ten.perature and upper shelf neutron shield tank (NST) material (A212-II) has energy level between the materials. However, the been characteriicd. Irradiation increases the materials exhibited essentially the same radiation Charpy t ransition temperatu re (CI'O by 23-28'C cmbrittlement sensitivity (both orientations),
(41-50"F) and decreases the upper-shelf energy.
proving that the anomaly is not due to a basic The shift in CITis not as severe as that observed difference in material irradiation resistances.
in high flux isotope reactor (HFIR) surveillance Possible causes of the original anomaly and the samples. Ilowever, the actual value of the CITis significance to NRC Regulatory Guide 1.99 are higher than that for the 11 FIR data. The increase discussed.
in yield stress is 51 MPa (7.4 ksi), which is compa-rable to IIFIR data. The NST material is weaker Iskander, S. K., et al., Oak Ridge National Laborato-in 1he transverse than in the longitudinal orienta-ry, "Results of Crack Arrest 1csts on Two Irradiated tion. Some effects of position across the thickness High Copper Welds," NUREG/CR-5584, U.S. Nu-of the wall are also observed; the CIT shift is clear Regulatory Commission, Washington, D.C.,
slightly greater for specimens from the inner re-December 1990.
gion of the wall. Annealing studies indicate com-
'Ihe objective of this study was to determine the plcte recovery from embrittlement after 1 h at effect of neutron irradiation on the shift and i
400 C(752*F). Although the wcld metalis sigmf-shape of the lower-bound curve to crack-arrest l
icantly tougher than the base metal, the shifts in data. Two submerged-arc welds with copper con-C1'I are comparable. The shifts m C1'I for the tents of 0.23 and 0.31 wt % were commercially Shippingport NST are consistent with the test and fabricated in 220-mm-thick plate. Crack-arrest Army reactor data for arradiations at < 232 C specimens fabricated from these welds were irra-(<450 F) and show very good agreement with diated at a nominal temperature of 288'C to an the results for liFIR A212-Il steel arradiated in average fluence of 1.9 x 1018 neutrons /cm2 (E >
the Oak Ridge Research Reactor (ORR). The 1 MeV). Evaluation of the results shows that the effects of irradiation temperature, fluence rate, neutron-irradiation-induced crack arrest tough-and neutron flux spectrum are discussed. Th ness temperature shift is about the sarne as the results mdicate that fluence rate has no effect on Charpy V-notch impact-temperature shift at the i
radiation embrittlement at rates as low as 2 x 108 41-J cncrgy level.The shape of the lower bound n/cm2/s and at the low operating temperatures of curves (for the range of test temperatures cov-the Shippingport NST, i.e., 55'C (130 F). This cred) did not seem to have been altered by irradi-suggests that the accelerated embrittlement of ation compared to those of the ASME K, curve.
t HFIR surveillance samples is most likely due to the relatively higher proportion of thermal neu-Kampmann, R., Institute For Materials Research, et trons in the liFIR spectrum compared to that for al.,"SANS Investigation of Low Alloy Steels in Neu-the test reactors.
tron Irradiated, Annealed, and Reirradiated Condi-tions," NUREG/CR-5926, U.S. Nuclear Regulatory Ilawthorne, J. R., Materials Engineering Associates, Commission, Washington, D.C., February 1993.
Inc., " Accelerated Irradiation 1bst of Gundrcmmin-gen Reactor Vessel Trepan Material," NUREG/
Small Angle Neutron Scattering (SANS) experi-CR-5891, U.S. Nuclear Regulatory Commission, ments were made on several low alioy steels and Washington, D.C., August 1992, submerged-arc wclds prototypic of nuclear reac-tor vessel construction. The objective was the Initial mechanical properties tests of beltline Ma-characterization of radiation-cnhanced and/or ra-terial trepanned from the decommissioned diation-induced precipitation contributing to me-l KRil-A pressure vessel and archive material irra-chanical property changes observed in tensile and diated in the UllR test reactor revealed a major notch ductility tests of the materials.The materi-anomaly in relative radiation cmbrittlement sen-als were irradiated in the UllR1bst Reactor un-sitivny.Poorcorrespondenceof materialbehavior dercloselycontrolled conditions. Aportion of the i
NUREG-1426 22 l
I
1 Compilation of Reports-1991-1993 j
samples were examined in the 288'C irradiated
'Ihe Fifth Irradiation Series in the IIcavy-Section (I) condition; others were examined in the postir-Steel Irradiation Program obtained a statistically radiation annealed (IA) condition and in the significant fmeture toughness data base on two 288'C reirradiated (I AR) condition. Experimen-high-copper (0.23 and 0.31 wt %) submerged-are i
tal variables included material composition (pri-welds to determine the shift and shape of the Kge marily %Cu, %P, %Ni content), postirradiation curve as a consequence of irradiation. Compact annealing temperature (454*C and 399"C) reir-specimens with thicknesses to 101.6 mm (4 in.)in radiation fluence level, and neutron-fluence rate the irradiated condition and 203.2 mm (8 in.)in
(-0.08,0.7, and 9 x 1012 n/cm2-s ', E > 1 hicV).
the unirradiated condition were tested, in addi-
'lhe apparent influence of the described variables tion to Charpy impact, tensile, and drop-weight on the size, number density, and composition of specimens. Irradiations were conducted at a nom-copper-rich p.ecipitates was the primary focus of inal temperature of 288'C and an average flu-the SANS analyses. SANS observations are re-ence of 1.5 x 10'8 n/cm2 (E > 1 MeV). The lated to measured notch ductility and tensile Charpy 41-J temperature shifts are about the property changes, with a view toward mechanistic same as the correspcmding drop-weight NDT explanation of the observed mechanical property temperature shifts. 'Ihe irradiated welds exhib-trends for I, IA, and IAR conditions.
ited substantial numbers of cleavage pop-ins.
Mean curve fits using two-parameter (with fixed Nanstad. R. K. and Ilerggren, R. G., Oak Ridge Na-intercept) nonlinear and linearized exponential n
tional Laboratory," Irradiation Effects on Charpy Im-regression analyses revealed that the fracture pact and Tensile Properties of Low Upper Shelf Welds, HSSI Series 2 and 3,, NUREG/CR-5696, toughness 100 Mlbd shifts exceeded the U.S. Nuclear Regulatory Commission, Washington, Charpy 41-J shifts for both welds. Analyses of D.C., August 1991.
curve shape changes indicated decreases in the slopes of the fracture toughness curves, especially
'Ihe objective of the Second and Third Irradiation for the higher copper weld. Weibull analyses were Series was to investigate the effects of irradiation performed to investigate development of lower on the ductile fracture toughness of seven com-bound curves to the data, including the use of a mercially fabricated, low upper-shelf welds. All variable Kmin parameter which affects the curve seven submerged-arc welds were fabricated with shape.
copper-coated wire and Linde 80 flux and had average bulk-copper contents from 0.21100.42%
Nanstad, R. K., et al., Oak Ridge National Labora-with nickel levels of about 0.6%. In addition to the tory, " Irradiation Effects on Fracture 'Ibughness of fracture toughness specimens which were irra.
Two liigh Copper Submerged-Arc Welds, llSSI Sc-diated at nominally 288'C, Charpy V-notch and ries 5, Appendices E and F," NUREG/CR-5913, Vol.
tensile specimens were included in the capsules at 2, U.S. Nuclear Regulatory Commission, Washing-available locations which were subject to wide ton, D.C., October 1992.
variations in irradiation temperature and fluence.
Ihis report presents analyses of the Charpy im-The Fifth Irradiation Series in the Heasy-Section Steel Irradiation Program obtained a statistically pact and tensile test data. Analyses revealed a significant fracture toughness data base on two dependence of yield strength on irradiation tem-perature of -1.1 MPa/*C, while the Charpy im-high-copper (0.23 and 0.31 wt %) submerged-are pact energy dependencies were about -0.5*C/*C welds to determine the shift and shape of the K el for transition temperature shift and -0.06 J/*C curve as a consequence of irradiation. Compact for upper-shelf decrease. After adjustment to an specimens with thicknesses to 101.6 mm (4 in.)in irradiation temperature of 288*C and normahza-the irradiated condition and 203.2 mm (8 in.)in the unirradiated condition were tested, in addi-tion to a fluence of 8 x 10ie neutrons /cm2(E > 1 McV), the Charpy transition temperature shifts tion to Charpy impact tensile, and drop-weight specimens. Irradiations were conducted at a nom-ranged from 59 to 123*C wmle the upper-shelf energies ranged from 58 to 79 J.
inal temperature of 288'C and an average flu-ence of 1.5 x 101o /cm2(E > 1 McV).The Charpy n
Nanstad, R. K., et al., Oak Ridge National Labora-tengatm s% am akut ty same as th tory, " Irradiation Effects on Fracture.Ibughness of correspondmg drop-weight ND'I temperature shifts. The irradiated welds exhibited substantial
,Iko High-Copper Submerged-Are Welds, IISSI Se; numbers of cleavage pop-ins. Mean curve fits us-ries 5, Main Report and Appendices A,11, C, and D, NUREG/CR-5913. Vol.1. U.S. Nuclear Regulatory ing two-parameter (with fixed intercept) nonlin-ear and linearized exponential regression Commission Washington, D.C., October 1992.
analyses revealed that the fracture toughness 100 MPad shifts exceeded the Charpy 41-J shifts 23 NUREG-1425
Compilation of Reports-1991-1993 for both welds. Analyses of curve shape changes such as void swelling and irradiation creep. Ile-
)
indicated decreases in the slopes of the frac.ure fore applying that theory to the much lower tem-toughness curves, especially for the higher copper perature and dose regimes characteristic of light weld. Weibull analyses were performed to investi.
water reactor pressure vessels and support struc-i gate development of lower bound curves to the tures, it is necessary to examine the assumptions data, including the use of a variable Kmin parame-made in formulating the theory. 'Re major sim-ter which affects the curve shape.
plifying assumption that has commonly been made is that the interstitial and vacancy concen-Stallmann, E W., et al., Oak Ridge National Labora-trations reach a quasi-steady state condition rap-tory,"PR EDil Power Reactor Embrittlement Data idly enough that the steady state concentrations llase, Version 1 Program Description," NUREG/
can be used in calculating the observable radi-FR-4816, U.S. Nuctcar Regulatory Commission, ation effects. The results presented here indicate Washington, D.C., July 1991.
that the assumption of steady state point defect concen trations it 'ot valid for temperat urcs much Data concerning radiation embrittlement of pres-below the light water reactor pressure vessel op-
< ure vessel stects in commercial power reactors crating temperature of about 288'C. At lower have been collected from available surveillance temperatures, the time required for the point reports.The purposc of this NRC-sponsored pro-defect concentrations to reach steady state can gram is to provide the technical bases for volun-execed an operating reactor's lifetime. Even at tary consensus standards, regulatory guides, 288'C, the point defect transient time can be long standard review plans, and codes. The data can enough to influence the interpretation of irradi-also be used for the exploration and verification of ation experiments done in materials test reactors embrittlement prediction models. The data files at accelerated damage rates. Based on the in-are given in dil ASE Ill PLUS format and can be sights obtained with the simple models of point accessed with any personal computer using the defect evolution, a more detailed model was de-DOS operating system. Menu-driven software is veloped that incorporates an explicit description provided for easy access to the data including of point defect clustering.1hese clusters are po-curve fitting and plotting facilitics.This software tentially responsible for the fraction of the radi-has drastically reduced the time and effort for ation-induced hardening that is attributed to the data processing and evaluation compared to pre-so-called " matrix defect." The model considers vious Ata bases.The current version of the Power both interstitial and vacancy clustering. The for-Reactor Embrittlement Data Base (PR-EDil) mer are treated as Frank loops while thc latter are lists the test results of 117 base materials (plates treated as microvoids. 'lhe point defect clusters and forgings), 85 wclds, and 88 heat affected-can be formed either directly in the displacement zonc materials that were irradiated in 241 cap-cascade or by diffusive encounters between free sules of 82 reactors. Many capsules also contained point defects.The results of molecular dynamics correlation materials (standard reference materi-simulation studies are used to provide guidance als, SRMs) from the ASTM plate and two llSST for the clustering parameters.1he hardening due plates (01 and 02). Material from the ilumboldt to point defect clusters was calculated using a Bay reactor was used as an SRM for some Gencr-simple dislocation barrier model. The results in-al Electric reactors.Tcc Electric Power Resce %
dicate that both interstitial and vacancy clusters Institute (EPRI), reacor vendors, and utilit.
can give rise to significant hardening.The relative have provided back-up asality assurance checks importance of cach cluster type is shown to be a of the PR-EI18.
function of irradiation temperature and displace-ment rate.
Stoller, R.
E., Oak Ridge National Laboratory, "Modeling the Influence of Irradiation Temperature Reactor Pressure Vessel Integrity and Displacement Rate on Radiation Induced liard-Assessments ening in Ferritic Steels," NUREG/CR-5859, U.S.
Nuclear Regulatory Commission, Washington, D.C.,
Cheverton, R. D., Oak Ridge National Laboratory, et July 1992.
al., " Review of Reactor Pressure Vessel Evaluation Report for Yankee Rowc Nuclear Power Station,"
.Ihc m. fluence ofirradiation temperature and dis-NUREG/CR-5799, U.S. Nuclear Regulatory Com-placement rate have been investigated usmg a mission, Washington, D.C., March 1992.
model based on the reaction rate theory descrip-tion of radiation damage. This theory was devel-The Yankee Atomic Electric Company (YAEC) ope.i primarily for the investigation of relatively has performed an Integrated Pressuri7ed Ther-highnemperature, high-dose radiation effects mal Shock (llTS)-type evaluation of the Yankee NUREG-1426 24
a Compilation of Reports-1991-1993 Rowe reactor pressure vessel in accordance with Reactor Pressure Vessel," NUREG/CR-5782, U.S.
the l'rS Rule (10 CFR 50.61)and U.S. Regulatory Nuclear Regulatory Comraission, Washington, D.C.,
Guide 1.154.The Oak Ridge National Laboratory August 1993.
(ORNL) reviewed the YAEC document and per-The Nuc! car Regulatory Commission (NRC) formed an mdependent probabilistic fracture-mechames analysis. The review included a requested Oak Ridge National Laboratory (ORNL) comparison of the Pacific Northwest Laboratory to perform a pressurized-thermal-shock (l'IS) pro-and the ORNL probabilistic fracture-mechames babilistic fracture mechanics (PFM) sensitivity anal-codes (VISA-Il and OCA-P, respectively). The ys s for the Yankee Rowe reactor pressure vessel, review identified minor errors and one signtficant for the fluences corresponding to the end of operat-I difference m philosophy. Also, the two codes ing cycle 21, using a specific small-break-loss-of-have a few dissimilar peripheral features. Aside coolant transient as the loading condition. Regions from these differences, VISA-Il and OCA-1 are of the vessel with distinguishing features were tobe very similar and with errors corrected and when treated individually-upper axial weld, k)wer axial adjusted for the difference in the treatment of weld, circumferential weld, upper plate snot welds, fracture toughness distribution through the wall, upper plate regions between the spot wc!ds, lower yield essentially the same value of the conditional plate spot welds, and the lower plate regions be-probability of failure. Ihc ORNL andependent tween the spot welds. De fracture analysis meth-evaluation indicated KrNIrr values considerably ods used in the analysis of through-clad surface greater than those corresponding to the l'rS-flaws were those contained in the established Rule screening criteria and a frequency of failure OCA-P computer code, which was developed dur-substantially greater than that corresponding to ing the Integrated Pressurized %crmal Shock the ' primary acceptance criterion"in U.S. Regu-
@'IS) Program. De NRC request specified thst latory Guide 1.154. ' lime constraints, however' the OCA-P code be enhanced for this study to also prevented as rigorous a treatment as the situatmn calculate the conditional probabilities of failure for deserves. Ihus, these results are very prehmi-subclad flaws and embedded flaws.'ne results of
""'Y' this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative Dickson, T. l_ Oak Ridge National Laboratory, influence f a number of key input paraineters in a
weI ana an at h an
" Generic Analysis for Evaluation of Low Charpy Upper-shelf Energy Effects on Safety Margins be used for readily determining the probability of Against Fracture r,I Reactor Pressure Vessel Materi-vml I Dum once a mom aaurate mdcaton d v
cm ment kcomes avanah M m als," NUREG/CR-6023, U.S. Nuclear Regulatory Commission, Washington, D.C., July 1993.
p rt is designated as HSST report No.117.
Appendix G to 10 CFR Part 50 requires that reac-Risk-Hased Inspection tor pressure vessel beltline materials maintain Charpy upper-shelf energies of no less than 50 American Society of Mechanical Engineers
- Risk ft-lb during the plant operating life, unless it is llased Inspection - Development of Guidelines.
demonstrated m a manner approved by the Nu-General Document," NUREG/GR-0005, Vol.1, clear Regulatory Commission (NRC), that lower U.S. Nuclear Regulatory Commission, Washington, valucs of Charpy upper-shelf energy provide mar-1).C., February 1992.
gins of safety against fracture equivalent to those Inservice inspection can play a significant role in in Appendix G to Section XI of the ASME Code.
minimizing equipment and structural failures.
Analyses based on acceptance criteria and analy-For many industrial applications, requirements sis methods adopted in the ASME Code Case for inservice inspection are based upon prior ex-N-512 are described herein. Additional informa-perience or engineering judgment, or are nonex-tion on material properties was provided by the istent. Most requirements or guidelines for;hese NRC, Office of Nuclear Regulatory Research, inspections are based on engineers' qualitative Materials Engineering Ilranch. Rese cases, spe-judgment, and only implicitly take into account cified by the 10RC, represent generic applications the probability of failure of a component under its to boiling water reactor and pressurized water operation and hiading conditions, and the conse-i reactor vessch. This report is designated as HSST quence of such failure,if it occurs.nis document Report No.140.
recommends appropriate methods for establish-1 ing a risk-based inspection program for any facil-Dickson,T. L, et al., Oak Ridge National Laboratory, ity or structural system. De process involves four
" Pressurized Thermal Shock Probabilistic Fracture major steps: defining the system; performing a Mechanics Sensitivity Analysis for Yankee Rowe qualitative risk assessment: using this to do a i
25 NUREG-1426
Compilation of Reports-1991-1993 quantitative risk analysis; and deve:oping an in-ment temperature range. The major variables in-spection program for components and structural vestigated included: (1) heating rate; (2) peak elements using probabilistic engineering meth-temperature; (3) holding time at peak temperature; ods. A companion document will detail specific and (4) cooling mte. Change in sensitization was risk-based techniques for the inspection of com-tracked using the electrochemical potentiokinetic ponents of LWR nuclear power plants, applying reactivation (EPR) test. Continuous heating / cool-methodology set out in Volume 1.
ing cyc!cs were performed using a furnace or using a thermal cycle simulation machine (Gleeble). Sensi-American Society of Mechanical Engineers, " Risk tization was found to increase with increasing peak liased Inspection - Development of Guidelines.
temperature until a " critical" peak temperature was Light Water Reactor (LWR) Nuclear Power Plant reached. Sensitization was very low for all samples Compments," NUREG/GR4X)05, Vol. 2, Part 1, heated above this critical peak temperature. 'the U.S. Nuclear Regulatory Commission, Washington, criticd peak temperature was 900*C for high-D.C., July 1993.
carbon (0.06wt%) 3M and varied from 950 to Effective inservice inspection progmms can play a 1000*C for high-carbon (0.06 wt%) 316 SS. Sensiti-z ti n increased with decreanng cooling rate and significant role in minimizing equipment and struc-tural failures. Most of the current inservice inspec-appeared to decrease with increasing heating rate.
,Ihc slowest heating rate used was equal to the tion programs for light water reactor (LWR) nuc! car power plant components are based on ex-f sist cooling rate tested. Results arc discussed m perience and engineers' qualitative judgment.
terms of gmm boundary chromium mrbide nuc!c-
'these progams include only an implicit consider-ation and preciptation, and chromium depletion.
ation of risk, which combines the protubility of fail-Atteridge, D. G., et al., Oregon G raduate Institute of ure of a compment under its operation and hiading Science & lbchnology," Measurement and Modeling co:!ditions and the consequences of such failure, if it of Sensitization Development in Stainless Steels as a occurs. lhis document recommends appropriate Function of Thermomechanical Processing,"
methods for establishing a risk-based inspection NUREG/GR4XX)4, U.S. Nuclear Regulatory Com-program for LWR nuclear power plant compo-mission, Washington, D.C., November 1992.
nents. *Ihc process has been built from a general methodology (Volume 1)and has been expanded to An analytical model has been developed for pre-involve five major steps: defining the system; eva-dicting thermomechanic:d effects on the develop-luating qualitative risk assessment results; using ment of grain boundary chromium depletion in this and information from plant probabilistic risk austenitic stainless steel as a first step in predict-assessments to perform a quantitative risk analysis; ing intergranular stress corrosion cracking sus-selecting target failure probabilitics; and developing ceptibility. Model development and validation is an inspection program for components using eco-based on sensitization development analysis of nomic decision analysis and structural reliability over 30'I}pc 316 and 304 stainicss stecl heats.The assessment methods. Included: extensive bibliogra-data base included analysis of deformation effects phy. Companion Volume 2 - Ibrt 2 document will on resultant sensiti7ation development. Continu-recommend risk-based inspection program for con-ous cooling sensitization behavior is examined sideration by Section XI of the ASME 11 oiler and and modeled with and without strain. Gas tung.
Pressure Vessel Code, sten are girth pipe weldments are also character-i7ed by experimental measurements of heat affected zone temperatures, strains and sensiti-Stam. less Steel Sens.t.i izat. ion zation during/after cach pass; pass by pass ther-ma to&s are aho pre &c h mWel h tkn usd to assus ppe chem %e changs on Atteridge, D. G., and Cedeno, C. A., Oregon Gradu-ate Institute of Science & 'Ibchnology," Continuous chromium depletion changes.
Cooling " thermal Cycle Effects on Sensitization in Stainless Steel" NUREG/GR-0002, U.S. Nuclear llruemmer, S. M., and Atteridge, D. G., Oregon Regulatory Commission. Washington, D.C., Septem-Graduate Institute of Science & 'Ibchnology,"Quan-ber 1991.
titative Measurement and Modeling of Sensitization Development in Stainless Steel," NUREG/
Work for this study was directed towards quantify-GR4X)01, U.S. Nuclear Regulatory Commission, ing sensiti7ation development (defined as gram Washington, D.C., September 1992 boundary chromium depletion)in high arbon'lype 3N and 316 stainless steel (SS) subjected to liocar
'1he state-of-the art to quantitatively measure and heating to a given peak temperature folh>wed by model sensitization development in austenitie stain-linear cooling through the sensitization develop-less stects is assessed and critically analyzed. A NUREG-1426 26
Compilation of Reports-1991-1993 modeling capabdity is evolved and validated using a quent continuous cooling. Strain recovery at cle-diverse experimental data base. Quantitative pre-vated temperatures played an important role in dictions are demonstrated for simpic and complex reducing t he effectiveness of prior deformation in thern.al and thermomechanical treatments. Com-accelerating rensitization kinetics. Due to the ef-mercial stainless steel heats ranging from high-car-fects of recovery, in certain cases, prior strain lxm 1}pe 3(4 and 316 to low-carbon 1}pe 3(ML and values of 20% were only as effective as 10% in 316L have been examined including many heats increasing the rate of sensitization development.
which correspond to extm-low-cad >on, nuclear-Limited transgranular carbide precipitation was grade compositions. Within certain limits the elec-observed in 20% prior strain samples but was not trochemical potentiokinetic reactivation (EPR) test a significant factor in the present work.1he was found to give accurate and reproducible mea-SSDOS model consistently over predicted DOS surements of the degree of sensitization (DOS)in development regardless of material condition.
'l}pe 3(4 and 316 stainless steels. EPR test results are used to develop the quantitative data base and evolyckdidate the quantitative modeling capabili-ty.1his thesis represents a first step to evolve meth-Thermal Aging (Cast Stainless Stcel) ods for the quantitative assessment of structural rehabihty in stainicss steel compments and weld-ments. Assessments will be based on compment-Chopra, O. K., Argonne National Laboratory, "Esti-specific information ccmcerning material mation of Fracture 1bughness of Cast Stainless Stects characteristics, fabrication history and service expo-During'lhermal Aging in LWR Systems," NUREG/
sure. Methods will enable fabrication (e.g., weldmg CR-4513, U.S. Nuclear Regulatory Commission, and repair weldmg) procedures and material aging Washington, D.C., June 1991, effects to be evaluated and ensure adequat" crack-ing resistance during the senice lifetime of reactor compments.1his work is being conducted by the A procedure and cortclations are presented for Oregon Gmduate Institute with interactive input predicting the change in fracture toughness of from personnel at Ihcific Northwest Laboratory.
cast stainless stect comp <ments due to thermal aging during senice in light water reactors (LWRs)at 280-330'C (535-625 F).1hc fracture toughness J-R curve and Charpy-impact energy Simmons, J. W., Oregon Graduate Institute of Sci-of aged cast stainicss steels are estimated from ence & Technology, *Effect of Prior Deformation on known material information. Fracture toughness Sensitization Development in Stainless Stect During of a specific cast..tainless steel is estimated from e
Continuous Cooling." NUREG/GR-4XX13. U.S. Nu-the extent and kinetics of thermal embrittlement.
clear Regulatory Commission. Washington, D.C.,
The extent of thermal embrittlement is charac-September 1991.
terized by the room-temperature " normalized" Charpy-impact energy. A orrelation for the ex-tent of embrittlement at " saturation," i.e., the liigh-cadxm 1}pc 316 stainless steel (SS) speci-minimum impact energy that would be achieved mens were subjected to linear continuous cooling for the material after long-term aging, is given in in a computer-controlled Glechte thermal simu-terms or a material parameter 4, which is deter.
lator, ne degree of sensitization (DOS) was mined from the chemical composition. The frac-quantitatively measured using the clectrochemi-ture toughness J-R curve for the material is then cal potent iokinetic reactivation (EPR) t est. Sensi-obtained from correlations between room-tem-tization values for the thermal cycles emph)yed in perature Charpy-impact energy and fracture the investigation were predicted using Ilruem-toughness parameters. Fracture toughness as a mer's SSDOS sensitization prediction model.
function of time and temperature of reactor scr-Prior deformation significantly cnhanced the rate vice is estimated from the kinetics of thermal of DOS development in the'l}pe 316 SS material.
embrittlement, which is determined from chemi-The DOS increased with increasing amounts of cal composition. A cornmon " lower-bound" J-R prior strain and decreasing cooling rates. Sensiti-curve for cast stainless stects with unknown zation respmse was also sensitive to peak cycle chemical composition is also defined for a given temperatures. Continuous cooling sensitization material specification, ferrite content, and tem-development occurred primarily in the critical perature. Examples for estimating impact temperature range between about 900 and strength and fracture toughness of cast stainless 750*C. Peak cycle temperatures above 1000'C steel components during reactor senice are de-retarded sensitiration development during subsc-scribed.
27 NUREG-1426
t Compilation of Reports-1991-1993 l
Chopra, O. K. and Chung, i1. M., Argonne National stress of the steels from data for the kinetics or i
Laboratory,"1 mg/Ibrm Embrittlement of Cast Du-thermal embrittlement.
plex Stainless Stects in LWR Systems," NUREG/
CR-4744, Vol. 4, No.1. U.S. Nuclear Regulatory Chopra, O. K. and Ilush, L Y., Argonne National Commission, Washington, D.C., May 1991.
laboratory,"lamgJIerm Embrittlement of Cast Du-plex Stainless Steels in LWR Systems. Semiannual His progress report summarizes work performed Report for October 1989 - March 1990," NUREG/
l by Argonne National Laboratory on long-term CR--4744, Vol. 5, No.1, U.S. Nuclear Regulatory embrittlement of cast duplex stainless stects in Commission, Washington, D.C., July 1991.
LWR systems during the six months from Octo-This progress report summarizes work performed ber 1988 to March 1989. Charpy-impact data are by ANL on long-term thermal embritt!cment of presented for several heats of cast stainless steel cast daplex stainless steels in LWR systems dur-aged at temperatures between 320 and 450*C for ing ite six months from October 1989 to March times up to 30,000 h. Thermal aging decreases 3990. De results from Charpy-impact tests and impact energy and shifts transition curves to high-miciohardnessmeasurementsof theferritephase er temperatures. A saturation effect is observed for several heats of cast stainless steel aged up to for room-temperature impact energy and up-30,000 h at 290-400*C are analyzed to establish per-shelf energy. Charpy data are analyzed to the kinetics of thermal embrittlement. Correla-obtain the activation energy of the kinetics of tions are presented for predicting the extent and embrittlement.nc results suggest that the acti-kinetics of thermal embrittlement of cast stain-vation energy of embrittlement is not constant in less stects from material information that can be i
the temperature range of 290-400*C, but in-determined from the certified material test re-creases as temperature decreases. A correlation cord. The extent of embrittlement is character-is presented for estimating the extent of em-ized by the room-temperature " normalized" brittlement of cast stainless steels from known Charpy-impact energy. Based on the information material parameters. The degradation in me-available, two methods are presented for estimat-chanical properties can be reversed by annealing ing the extent of embrittlement at " saturation,"
the embrittled material for 1 h at 550*C and then i.e., the minimum impact energy that would be water quenchmg.
achieved for the material after long-term aging.
The first method utilizes only the chemical com-Chopra. O. K., et al., Argonne National Laboratory, position of the steci.The second method is used "Long/lbrm Embrittlement of Cast Duplex Stainless when metallographic mformation on the ferrite Steels in LWR Systems, Semitmnual Report for April morphology, i.e., ferrite content and mean ferrit c
- September 1989," NUREG/CR-4744, Vol. 4, No. 2, spacing of the stcel, is also available. The change U.S. Nuclear Regulatory Commission, Washington, m Charpy-impact energy as a function of time and D.C., June 1991.
temperature of reactor service is then estimated from the extent of embrittlement at saturation and from the correlations describing the kinetics This progress report summarizes work performed of embrittlement, which is expressed in terms of by Argonne National Laboratory on long term the chemical composition and agmg behavior of thermal embrittlem(nt of cast duplex stainicss the steel at 400*L,.
stects in LWR systems during the six me-dis from April to September 1989. Tensile r.nd fracture Chopra, O.
K.,
Argonne National Laboratory, toughness data are presented for sc.cral heats of
- l.nng!Ibrm Embrittlement of Cast Duplex Stainless cast stainless steel that were aged up to 30,000 h at Stects in LWR Systems. Semiannual Report for April temperatures of 290-450'C.De results mdicale
- September 1990." N UREG/CR-4744, Vol. 5, No. 2, that thermal agmg mercases the tensile stress and U.S. Nuclear Regulatory Commission, Washington, i
decreases the fracture toughness of the materials.
D.C., July 1991.
t in general, C-3 steels are the least sensitive to thermal aging embrittiement and C-8M steels His progress report summarizes work performed t
are the most sensitive.nc increase in flow stress by Argonne National Laboratory on long-term em-of fully aged cast stainicss steels is "10% for brittlement of cast duplex stainless steels in LWR CF-3 stects and =20% for CF-8 and CF-SM systems during the six months from April-Septem-stects. The fracture toughness J c and average her 1990. A procedure and correlations are pres-3 tearing modulus for heats that are sensitive to ented for predicting fmeture toughness J.R curves thermal aging (e.g., CF-SM steels) are as low as and impact strength of aged cast stainless stects a
~90 kJ/m2 and =60, respectively. Correlations from known materi:d information. Fracture tough-are presented for estimating the increase in flow ness of a specific cast stainless steel is estimated i
NUREG-1426 28
Compilation of Reports-1991-1993 l
i from the extent and kinetics of embrittlement. 'lhe April - September 1991," NUREG/CR-4744, Vol. 6, exterit of embrittlement is characterized by the No. 2, U.S. Nuclear Regulatory Commission, Wash-rmm-temperature Charpy-impact energy. A mrre-ington, D.C., November 1992.
lation for the extent of embrittlement at saturation is given in terms of a material parameter, phi, w hich This progress report summarizes work performed is determined from the ferrite morphology and/or by Argonne National Laboratory on long-term chemical composition. Charpy-tmpact energy as a embrittlement of cast duplex stainless steels in function of time and temperature of reactor semce LWR systems during the six months from April to is estimated from the kinctus of embrittlement, September 1991. A pmcedure and correlations yhich are determined from chemical composition.
are presented for predicting Charpy-impact ener-1he fracture toughness J R curse for the material is gy, tensile flow stress, fracture toughness J-R then obtamed from correlations between curve, and Jje of aged cast stainless steels from room-temperature Charpy-impact energy and frac-known material information.1he " saturation" ture tougimess pa Ameters. A lower-bound' J-lj impact strength and fracture toughness of a spe-curve for cast stan. ess stects w,th unknown chemi-cific cast stainless steel, i.e., the minimum value i
cal composition is.Jso defined for a given material that would be achieved for the material after specification and tempat'.re.. -chanical-proper-long-term service, is estimated from the chemical ty degradation suffered by cast stainlee steel com-composition of the steel. Mechanical properties ponents from the decommissioned Shippngport as a function of time and temperature of reactor reactor has been charactcrued.The results are used service are estimated from impact energy and to validate the correlations and benchmark the lab-flow stress of Ihe unaged matcrial and the kinetics oratory studies. Charpy-impact, tensile, and frac-of embrittlement, which are also determined ture toughness data for materials from the hot-leg from chemical composition. The J ei values are shutoff valve and cold-leg check valves and pump determined from the estimated J-R curve and volute are presented. 'lhe results mdicate a modest flow stress. Exampics of estimating mechanical degree of embrittlement.
properties of cast stainless steel components dur-ing reactor service are presented. A common
- Chopra, O.
K.,
Argonne National Laboratory, nd"J43 mm for cast stainless steels wu "leng. Term Embrittlement of Cast Duplex Stainless f unknown chemical composition = also defined Steels in LWR Systems. Semiannual Report for Octo-f r a given grade of steel, fernte content, and r
her 1WO - March 1991," NUREG/CR-4744, Vol. 6.
tempuature.
No.1, U.S. Nuclear Regulatory Commission, Washing.
ton, D.C., August 1992.
Chopra, O.
K., Argonne National Laboratory,
[
"Long-1brm Embrittlement of Cast Duplex Stainless This progress report summarizes work performed Steels in LWR Systems. Semiannual Report for Octo-i by Argonne National Laboratory on long-tem ber 1991 - March 1992," NUREG/CR-4744, Vol. 7
[
thermal embrittlement of ast duplex stainless No.1, U.S. Nuclear Regulatory Commission, Wash-stcels in LWR systems dunng the six months from ington, D.C., May 1993.
October 1990 to March 1991. Charpy-impact, ten.
This progress report summarizes work performed sile, and fracture toughness data are prerented by Argonne National Laboratory on long-term l
for several heats of cast stainless steel that were thermal embrittlement of cast duplex stainless aged up to 58,0(X) h at temperatures of steels in LWR systems during the six months from 290-400*C.The results indicate that thermal ag-October 1991 to March 1992. Charpy-impact, ing increases the tensile stress and decreases the tensile, and fracture toughness J-R curve data are fracture toughness of the materials. In general, presented for several heats of cast stainless steel CF-3 steels are the least sensitive to thermal that were aged 10,000-58,000 h at 290,320, and aging embrittlement and CF-SM steels are the 350*C. The results indicate that thermal aging most sensitive.The increase in flow stress of fully decreases the fracture toughness of cast stainless aged cast stainless steels is =10% for CF-3 steels stects. In general, CF-3 steels are the least sensi-and =20% for CF-8 and CF-SM steels.The frac-tive to thermal aging and CF-SM stects are the l
ture toughness J et and average tearing modulus most sensitive. The values of fracture toughness i
for heats that are sensitive to thermal aging (e.g.
Jgeand tearing modulus for CF-8M steels can be CF-SM steels) are as low as =90 kJ/m2 and =60, as low as =90 kJ/m2 and =60, respectively.'Ihe respectively.
fracture toughness data are consistent with the Charpy-impact results, i.e., unaged and aged Chopra, O.
K., Argeane National laboratory, steels that show low impact energy also exhibit "lengZltrm Embrittlemet of Cast Duplex Stainless lower fracture toughness. All steels reach a mini-Steels in LWR Systems. Semiannual Report for mum saturation fracture toughness after thermal i
29 NUREG-1426
Compilation of Itcports-1991-1993 aging; the time to reach saturation depends on the less stects from known material information.The aging temperature.The results also indicate that
" saturation" impact strength and fracture tough-low-strength cast stainless steels are generally ness of a specific cast stainless stect, i.e., the mini-insensitive to thermal aging.
mum value that would be achieved for the material after long-term service, is estimated Chopra, O.
K., Argonne National Laboratory, from the chemical composition of the steel. Me-
"long-Term Embrittlement of Cast Duplex Stainless chanical properties as a function of time and tem-Steels in LWII Systems. Semiannual lleport for perature of reactor service are estimated from April-September 1992." NUllEG/ Cit-4744, Vol. 7 mpact energy and flow stress ofIhe unaged mate-No. 2, U.S. Nuclear Itegulatory Commission, Wash-rial and the kinetics of embrittlement, which are ington, D.C., July 1993, also determined from chemical composition.The lci values are determined from the estimated J-It This progress repor t summarizes work performed by Argonne National Laboratory on long-term curve and flow stress. Examples of estimating me-thermal ernbrittlement of cast duplex stain! css chanical properties of cast stainless steel compo-stecis in LWil systems during the six months from nents during reactor service are presented. A April-September 1992. A procedure and correla-comrnon " lower-bound" J-Il curve for cast stain-tions are presented for predicting Charpy-impact less stects of unknown chemical composition is energy tensile flow stress, fracture toughness J.It also defined for a given grade of stecl. ferrite i
curves, tearing modulus and J ei of aged cast stain-content, and temperature.
1 F
i i
P NUllEG-1426 30
t#4C FORM 335 U.S. NUCLEAR HEGut4 TORY COMMISSION
- 1. REPORT NUMBER (249)
(Assigned by NRC. Add Vol..
I t&4CM 1102, Supp., Rev., and Addendum Num-32on 3202 BIBLIOGRAPHIC DATA SHEET b" * - " anr4 i
(Se. instructions on the re ws.)
- a. Tate ANo SUBmLE Vol. 2
- 3. DATE HEPORT PUBUSHED Compilation of Reports from Research Supported by the Materials Engineering MONTH YEAR I
Branch, Division of Engineering g
1991-1993
- b. AUTHOR (5)
- 6. TYPE OF REPORT Compiled by A. L Hiser, Jr.
"Ibchnical
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Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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U.S. Nuclear Regulatory Commission, and msHing address.)
Same as above.
l
- 10. SUPPLEMENT ARY NOl ES 11, ABSTRACT (200 words or less)
Since 1%5, the Materials Engineering Branch, Division of Engineering, of the Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, and its predecessors dating back to the Atomic Energy Commission (AEC),
i has sponsored research programs concerning the integrity of the primary system pressure boundary of light water reactors. The components of concern in these research programs have included the reactor pressure vessel (RPV),
i steam generators, and the piping. These research programs have covered a broad range of topics, including fracture mechanics analysis and experimental work for RPV and piping applications, inspection method development wl i
qualification, and evaluation of irradiation effects to RPV steels.
His report provides as mmplete a listing as practical of formal technical reports submitted to the NRC by the inves-tigators working on these research programs. This listing includes topical, final and progress reports, and is seg-mented by topic area. In many cases a report will cover several topics (such as in the case of progress reports of multi-faceted programs), but is listed under only one topic. Therefore, in searching for reports on a specific topic, l
other related topic areas should be checked also. He previous volume to this report covers the period 1965 - 1990.
i i
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- 13. AVAILABluTY STATEMENT Unlimited
- 14. SECURITY CLASSIFICATION reactor pressure vessels, piping, fracture mechanics, non-destructive examination, g,,,,,g radiation embrittlement, dosimetry, environmentally-assisted cracking, fatigue, steam generators, annealing, research reports Unclassified 4
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