WorldCat Identities

Shack, W. J.

Overview
Works: 64 works in 101 publications in 1 language and 4,605 library holdings
Roles: Author
Classifications: TK9166, 621.483
Publication Timeline
.
Most widely held works by W. J Shack
Fracture toughness and crack growth rates of irradiated austenitic stainless steels by O. K Chopra( )

2 editions published in 2003 in English and held by 340 WorldCat member libraries worldwide

Effects of alloy chemistry, cold work, and water chemistry on corrosion fatigue and stress corrosion cracking of nickel alloys and welds by O. K Chopra( )

2 editions published in 2001 in English and held by 315 WorldCat member libraries worldwide

Environmental effects on fatigue crack initiation in piping and pressure vessel steels by O. K Chopra( )

2 editions published in 2001 in English and held by 314 WorldCat member libraries worldwide

Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels by O. K Chopra( )

2 editions published in 1998 in English and held by 291 WorldCat member libraries worldwide

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments by O. K Chopra( )

2 editions published in 2008 in English and held by 267 WorldCat member libraries worldwide

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as H"2x 10²¹ n/cm² (E> 1 MeV) (H"3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated
Crack growth rates and metallographic examinations of alloy 600 and alloy 82/182 from field components and laboratory materials tested in PWR environments by B Alexandreanu( )

3 editions published in 2008 in English and held by 265 WorldCat member libraries worldwide

In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed
Mechanical properties of thermally aged cast stainless steels from Shippingport Reactor components by O. K Chopra( )

2 editions published in 1995 in English and held by 253 WorldCat member libraries worldwide

Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was -13 y at -281 C (538 F) for the hot-leg components and -264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot and crossover-leg elbows (CF-8M steel) after service of - 15 y and the KRB reactor pump cover plate (CF-8) after - 8 y of service
Effect of LWR coolant environments on the fatigue life of reactor materials by O. K Chopra( )

4 editions published between 2006 and 2007 in English and held by 235 WorldCat member libraries worldwide

Assessment of noise level for eddy current inspection of steam generator tubes by Sasan Bakhtiari( )

3 editions published in 2009 in English and held by 217 WorldCat member libraries worldwide

Crack growth rates of irradiated nickel alloy welds in a PWR environment by B Alexandreanu( )

1 edition published in 2006 in English and held by 213 WorldCat member libraries worldwide

In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of ≈5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking
Crack growth rates in a PWR environment of nickel alloys from the Davis-Besse and V.C. Summer power plants by B Alexandreanu( )

3 editions published in 2006 in English and held by 211 WorldCat member libraries worldwide

Review of the margins for ASME code fatigue design curve : effects of surface roughness and material variability by O. K Chopra( )

1 edition published in 2003 in English and held by 210 WorldCat member libraries worldwide

Effect of material heat treatment on fatigue crack initiation in austenitic stainless steels in LWR environments by O. K Chopra( )

1 edition published in 2005 in English and held by 210 WorldCat member libraries worldwide

The ASME Boiler and Pressure Vessel Code provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify design curves for applicable structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. The existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. Under certain environmental and loading conditions, fatigue lives of austenitic stainless steels (SSs) can be a factor of 20 lower in water than in air. This report presents experimental data on the effect of heat treatment on fatigue crack initiation in austenitic Type 304 SS in LWR coolant environments. A detailed metallographic examination of fatigue test specimens was performed to characterize the crack morphology and fracture morphology. The key material, loading, and environmental parameters and their effect on the fatigue life of these steels are also described. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic SSs as a function of material, loading, and environmental parameters. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented
Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals by H. M Chung( )

1 edition published in 2006 in English and held by 209 WorldCat member libraries worldwide

This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at (almost equal to)3 dpa is a good measure of IASCC susceptibility. At (almost equal to)1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At (almost equal to)3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to (almost equal to)3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain>0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to the behavior of their high-C counterparts. At S concentrations>0.002 wt.%, the deleterious effect of S is so dominant that a high concentration of C is not an important factor. A two-dimensional map was developed in which susceptibility or resistance to IASCC is shown as a function of bulk concentrations of S and C. Data reported in the literature are consistent with the map. The map is helpful to predict relative IASCC susceptibility of Types 304 and 316 steels. A similar but somewhat different map is helpful to predict IASCC behavior of Type 348 steels. Grain-boundary segregation of S was observed for BWR neutron absorber tubes irradiated to (almost equal to)3 dpa. On the basis of the results of the stress-corrosion-cracking tests and the microstructural characterization, a mechanistic IASCC model has been developed
Application of automated analysis software to eddy current inspection data from steam generator tube bundle mock-up by Sasan Bakhtiari( )

2 editions published in 2016 in English and held by 206 WorldCat member libraries worldwide

Behavior of PWR reactor coolant system components, other than steam generator tubes, under severe accident conditions : phase I final report by S Majumdar( Book )

1 edition published in 2003 in English and held by 103 WorldCat member libraries worldwide

Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents by S Majumdar( Book )

1 edition published in 2002 in English and held by 94 WorldCat member libraries worldwide

Radiation embrittlement of the neutron shield tank from the Shippingport reactor by O. K Chopra( Book )

3 editions published in 1991 in English and held by 93 WorldCat member libraries worldwide

Review of environmental effects on fatigue crack growth of austenitic stainless steels by W. J Shack( Book )

2 editions published in 1994 in English and held by 92 WorldCat member libraries worldwide

Assessment of thermal embrittlement of cast stainless steels by O. K Chopra( Book )

2 editions published in 1994 in English and held by 92 WorldCat member libraries worldwide

 
moreShow More Titles
fewerShow Fewer Titles
Audience Level
0
Audience Level
1
  Kids General Special  
Audience level: 0.50 (from 0.46 for Crack grow ... to 0.60 for Assessment ...)

Languages
English (40)