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Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems - Revison 1.

Description: This report presents a revision of the procedure and correlations presented earlier in NUREG/CR-4513, ANL-90/42 (June 1991) for predicting the change in mechanical properties of cast stainless steel components due to thermal aging during service in light water reactors at 280-330 C (535-625 F). The correlations presented in this report are based on an expanded data base and have been optimized with mechanical-property data on cast stainless steels aged up to {approx}58,000 h at 290-350 C (554-633 F). The correlations for estimating the change in tensile stress, including the Ramberg/Osgood parameters for strain hardening, are also described. The fracture toughness J-R curve, tensile stress, and Charpy-impact energy of aged cast stainless steels are estimated from known material information. Mechanical properties of a specific cast stainless steel are estimated from the extent and kinetics of thermal embrittlement. Embrittlement of cast stainless steels is characterized in terms of room-temperature Charpy-impact energy. The extent or degree of thermal embrittlement at 'saturation,' i.e., the minimum impact energy that can be achieved for a material after long-term aging, is determined from the chemical composition of the steel. Charpy-impact energy as a function of time and temperature of reactor service is estimated from the kinetics of thermal embrittlement, which are also determined from the chemical composition. The initial impact energy of the unaged steel is required for these estimations. Initial tensile flow stress is needed for estimating the flow stress of the aged material. The fracture toughness J-R curve for the material is then obtained by correlating room-temperature Charpy-impact energy with fracture toughness parameters. The values of JIC are determined from the estimated J-R curve and flow stress. A common 'predicted lower-bound' J-R curve for cast stainless steels of unknown chemical composition is also defined for a given grade of steel, range of ferrite content, and ...
Date: October 5, 1994
Creator: Chopra, O. K. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Mechanism of fatigue crack initiation in austenitic stainless steels in LWR environments.

Description: This paper examines the mechanism of fatigue crack initiation in austenitic stainless steels (SSs) in light water reactor (LWR) coolant environments. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, level of dissolved oxygen in water, and material heat treatment on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The influence of reactor coolant environments on the formation and growth of fatigue cracks in polished smooth SS specimens is discussed. Crack length as a function of fatigue cycles was determined in air and LWR environments. The results indicate that decreased fatigue lives of these steels are caused primarily by the effects of the environment on the growth of cracks <200 {micro}m and, to a lesser extent, on enhanced growth rates of longer cracks. A detailed metallographic examination of fatigue test specimens was performed to characterize the fracture morphology. Exploratory fatigue tests were conducted to enhance our understanding of the effects of surface micropits or minor differences in the surface oxide on fatigue crack initiation.
Date: March 27, 2002
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

Description: 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 {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 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.
Date: January 21, 2008
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. 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. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue {var_epsilon}-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.
Date: August 1, 2002
Creator: Chopra, O. K. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Methods for incorporating effects of LWR coolant environment into ASME code fatigue evaluations.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Appendix I to Section HI of the Code specifies design fatigue curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of LWR environments on the fatigue resistance of carbon and low-alloy steels and austenitic stainless steels (SSs). Under certain loading and environmental conditions, fatigue lives of carbon and low-alloy steels can be a factor of {approx}70 lower in an LWR environment than in air. These results raise the issue of whether the design fatigue curves in Section III are appropriate for the intended purpose. This paper presents the two methods that have been proposed for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations. The mechanisms of fatigue crack initiation in carbon and low-alloy steels and austenitic SSs in LWR environments are discussed.
Date: April 15, 1999
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Effects of LWR coolant environments on fatigue lives of austenitic stainless steels.

Description: Fatigue tests have been conducted on Types 304 and 316NG stainless steels to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on the fatigue lives of these steels. The results confirm significant decreases in fatigue life in water. Unlike the situation with ferritic steels, environmental effects on Types 304 and 316NG stainless steel are more pronounced in low-DO than in high-DO water. Experimental results have been compared with estimates of fatigue life based on a statistical model. The formation and growth of fatigue cracks in air and water environments are discussed.
Date: January 13, 1998
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Estimation of mechanical properties of cast stainless steels during thermal aging in LWR systems

Description: A procedure and correlations are presented for assessing thermal embrittlement and predicting mechanical properties of cast stainless steels under light water reactor operating conditions from known material information. The ``saturation`` fracture toughness of a cast stainless steel, i.e., the minimum value that would be achieved for the material after long-term service, is estimated from the chemical composition of the steel. Fracture properties as a function of time and temperature of service are estimated from the kinetics of embrittlement, which are also determined from chemical composition. The correlations successfully predict fracture toughness, Charpy-impact, and tensile properties of cast stainless steels from the Shippingport, Ringhals, and KRB reactors.
Date: March 1, 1995
Creator: Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Methods for incorporating the effects of LWR coolant environments in pressure vessel and piping fatigue evaluations.

Description: This paper summarizes the work performed at Argonne National Laboratory on the fatigue of piping and pressure vessel steels in the coolant environments of light water reactors. The existing fatigue strain vs. life ({var_epsilon}-N) data were evaluated to establish the effects of various material and loading variables, such as steel type, strain range, strain rate, temperature, and dissolved-oxygen level in water, on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for carbon and low-alloy steels and austenitic stainless steels as a function of material, loading, and environment variables. Case studies of fatigue failures in nuclear power plants are presented, and the contribution of environmental effects to crack initiation is discussed. Methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the possible conservatism in the existing fatigue design curves of the ASME Code.
Date: July 31, 2002
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

Description: 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.
Date: May 5, 2008
Creator: Alexandreanu, B.; Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. 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. This report provides an overview of the existing fatigue {var_epsilon}-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented.
Date: October 3, 2003
Creator: Chopra, O. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Estimation of fatigue strain-life curves for austenitic stainless steels in light water reactor environments.

Description: The ASME Boiler and Pressure Vessel Code design fatigue curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue lives of austenitic stainless steels (SSs) in light water reactor (LWR) environments. Unlike those of carbon and low-alloy steels, environmental effects on fatigue lives of SSs are more pronounced in low-dissolved-oxygen (low-DO) water than in high-DO water, This paper summarizes available fatigue strain vs. life data on the effects of various material and loading variables such as steel type, DO level, strain range, and strain rate on the fatigue lives of wrought and cast austenitic SSs. Statistical models for estimating the fatigue lives of these steels in LWR environments have been updated with a larger data base. The significance of the effect of environment on the current Code design curve has been evaluated.
Date: February 12, 1998
Creator: Chopra, O. K. & Smith, J. L.
Partner: UNT Libraries Government Documents Department

Fatigue crack initiation in carbon and low-alloy steels in light water reactor environments : mechanism and prediction.

Description: Section 111 of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. The effects of reactor coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be shorter than those in air by a factor of {approx}70. The crack initiation and crack growth characteristics of carbon and low-alloy steels in LWR environments are presented. Decreases in fatigue life of these steels in high-dissolved-oxygen water are caused primarily by the effect of environment on growth of short cracks < 100 {micro}m in depth. The material and loading parameters that influence fatigue life in LWR environments are defined. Fatigue life is decreased significantly when five conditions are satisfied simultaneously, viz., applied strain range, service temperature, dissolved oxygen in water, and S content in steel are above a threshold level, and loading strain rate is below a threshold value. Statistical models have been developed for estimating the fatigue life of these steels in LWR environments. The significance of the effect of environment on the current Code design curve is evaluated.
Date: January 27, 1998
Creator: Chopra, O. K. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Mechanical properties of thermally aged cast stainless steels from shippingport reactor components.

Description: 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 {approx}13 y at {approx}281 C (538 F) for the hot-leg components and {approx}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 JIC 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 {approx}15 y and the KRB reactor pump cover plate (CF-8) after {approx}8 y of service.
Date: June 7, 1995
Creator: Chopra, O. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Studies of aged cast stainless steel from the Shippingport reactor

Description: The mechanical properties of cast stainless steels from the Shippingport reactor have been characterized. Baseline properties for unaged materials were obtained from tests on either recovery-annealed material or material from a cooler region of the component. The materials exhibited modest decrease in impact energy and fracture toughness and a small increase in tensile strength. The fracture toughness J-R curve, J{sub IC} value, tensile flow stress, and Charpy-impact energy of the materials showed very good agreement with estimations based on accelerated laboratory aging studies. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy that would be achieved after long-term aging, were established from materials that were aged further in the laboratory at temperatures between 320 and 400{degree}C. The results showed very good agreement with estimates; the activation energies ranged from 125 to 250 kJ/mole and the minimum room-temperature impact energy was >75 J/cm{sup 2}. The estimated impact energy and fracture toughness J-R curve for materials from the Ringhals reactor hot and crossover-leg elbows are also presented.
Date: October 1991
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Crack initiation in smooth fatigue specimens of austenitic stainless steel in light water reactor environments.

Description: The fatigue design curves for structural materials specified in Section III of the ASME Boiler and Pressure Vessel Code are based on tests of smooth polished specimens at room temperature in air. The effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves; however, recent test data illustrate the detrimental effects of LWR coolant environments on the fatigue resistance of austenitic stainless steels (SSs). Certain loading and environmental conditions have led to test specimen fatigue lives that are significantly shorter than those obtained in air. Results of fatigue tests that examine the influence of reactor environments on crack initiation and crack growth of austenitic SSs are presented. Block loading was used to mark the fracture surface to determine crack length as a function of fatigue cycles in water environments, Crack lengths were measured by scanning electron microscopy. The mechanism for decreased fatigue life in LWR environments is discussed, and crack growth rates in the smooth fatigue specimens are compared with existing data from studies of crack growth rates.
Date: April 8, 1999
Creator: Chopra, O. K. & Smith, J. L.
Partner: UNT Libraries Government Documents Department

Statistical analysis of fatigue strain-life data for carbon and low-alloy steels

Description: The existing fatigue strain vs life (S-N) data, foreign and domestic, for carbon and low-alloy steels used in the construction of nuclear power plant components have been compiled and categorized according to material, loading, and environmental conditions. A statistical model has been developed for estimating the effects of the various test conditions on fatigue life. The results of a rigorous statistical analysis have been used to estimate the probability of initiating a fatigue crack. Data in the literature were reviewed to evaluate the effects of size, geometry, and surface finish of a component on its fatigue life. The fatigue S-N curves for components have been determined by applying design margins for size, geometry, and surface finish to crack initiation curves estimated from the model.
Date: March 1, 1995
Creator: Keisler, J. & Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Effect of thermal aging on mechanical properties of cast stainless steels

Description: A procedure and correlations are presented for predicting mechanical properties of cast stainless steels in service at temperatures <450{degrees}C from known material information. The ``saturation`` fracture properties of a cast stainless steel, i.e., the minimum values that would be achieved for the material after long-term service, are estimated from the chemical composition of the steel. Fracture properties as a function of time and temperature of service are estimated from the kinetics of embrittlement, which are also determined from chemical composition. The correlations successfully predict fracture toughness, Charpy-impact, and tensile properties of cast stainless steels from the Shippingport-, Ringhals-, and Gundremmingen-reactor components.
Date: March 1, 1995
Creator: Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Compatibility of ITER candidate materials with static gallium

Description: Corrosion tests have been conducted to determine the compatibility of gallium with candidate structural materials for the International Thermonuclear Experimental Reactor (ITER) first wall/blanket systems, e.g., Type 316 stainless steel (SS), Inconel 625, and Nb-5 Mo-1 Zr. The results indicate that Type 316 SS is least resistant to corrosion in static gallium and Nb-5 Mo-1 Zr alloy is most resistant. At 400 C, corrosion rates for Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy are {approx} 4.0, 0.5, and 0.03 mm/yr, respectively. Iron, nickel, and chromium react rapidly with gallium. Iron shows greater corrosion than nickel at 400 C ({ge} 88 and 18 mm/yr, respectively). The present study indicates that at temperatures up to 400 C, corrosion occurs primarily by dissolution and is accompanied by formation of metal/gallium intermetallic compounds. The growth of intermetallic compounds may control the overall rate of corrosion.
Date: September 1995
Creator: Luebbers, P. R. & Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Estimation of mechanical properties of cast stainless steels during thermal aging in LWR systems

Description: A procedure and correlations are presented for predicting Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of aged cast stainless steels from known material information. The ``saturation`` impact strength and fracture toughness of a specific cast stainless steel, i.e., the minimum value that would be achieved for the material after long-term service, is estimated from the chemical composition of the steel. Mechanical properties as a function of time and temperature of reactor service are estimated from impact energy and flow stress of the unaged material and the kinetics of embrittlement, which are also determined from chemical composition. The J{sub IC} values are determined from the estimated J-R curve and flow stress. Examples of estimating mechanical properties of cast stainless steel components during reactor service are presented. A common predicted lower-bound J-R curve for cast stainless steels of unknown chemical composition is also defined for a given grade of steel, ferrite content, and temperature.
Date: October 1991
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

Description: A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of {Delta}J and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values.
Date: April 10, 2000
Creator: Park, H. B. & Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Fracture toughness and crack growth rates of irradiated austenitic stainless steels.

Description: Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. Experimental data are presented on fracture toughness and crack growth rates (CGRs) of austenitic SSs irradiated to fluence levels up to 2.0 x 10{sup 21} n/cm{sup 2} (E &gt; 1 MeV) ({approx}3.0 dpa) at {approx}288 C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor (BWR) environments, and fracture toughness tests were conducted in air. Neutron irradiation at 288 C decreases the fracture toughness of the steels; the data from commercial heats fall within the scatter band for the data obtained at higher temperatures. In addition, the results indicate significant enhancement of CGRs of the irradiated steels in normal water chemistry BWR environment; the CGRs for irradiated steels are a factor of {approx}5 higher than the disposition curve proposed for sensitized austenitic SSs. The rates decreased by more than an order of magnitude in low-dissolved-oxygen BWR environment.
Date: June 25, 2003
Creator: Chopra, O. K.; Gruber, E. E. & Shack, W. J.
Partner: UNT Libraries Government Documents Department