Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds.

Reactor vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking (EAC). A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. The objective of this work is to evaluate and compare the resistance of Alloys 600 and 690 and their welds, such as Alloys 82, 182, 52, and 152, to EAC in simulated light water reactor environments. The existing crack growth rate (CGR) data for these alloys under cyclic and constant loads have been evaluated to establish the effects of alloy chemistry, cold work, and water chemistry. The experimental fatigue CGRs are compared with CGRs that would be expected in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions for significant environmental enhancement in growth rates. The existing stress corrosion cracking (SCC) data on Alloys 600 and 690 and Alloy 82, 182, and 52 welds have been compiled and analyzed to determine the influence of key parameters on growth rates in simulated PWR and BWR environments. The SCC data for these alloys have been evaluated with correlations developed by Scott and by Ford and Andresen

Report on thermal aging effects on tensile properties of ferritic-martensitic steels.

This report provides an update on the evaluation of thermal-aging induced degradation of tensile properties of advanced ferritic-martensitic steels. The report is the first deliverable (level 3) in FY11 (M3A11AN04030103), under the Work Package A-11AN040301, 'Advanced Alloy Testing' performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing tensile data on aged alloys and a mechanistic model, validated by experiments, with a predictive capability on long-term performance. The scope of work is to evaluate the effect of thermal aging on the tensile properties of advanced alloys such as ferritic-martensitic steels, mod.9Cr-1Mo, NF616, and advanced austenitic stainless steel, HT-UPS. The aging experiments have been conducted over a temperature of 550-750 C for various time periods to simulate the microstructural changes in the alloys as a function of time at temperature. In addition, a mechanistic model based on thermodynamics and kinetics has been used to address the changes in microstructure of the alloys as a function of time and temperature, which is developed in the companion work package at ANL. The focus of this project is advanced alloy testing and understanding the effects of long-term thermal aging on the tensile properties. Advanced materials examined in this project include ferritic-martensitic steels mod.9Cr-1Mo and NF616, and austenitic steel, HT-UPS. The report summarizes the tensile testing results of thermally-aged mod.9Cr-1Mo, NF616 H1 and NF616 H2 ferritic-martensitic steels. NF616 H1 and NF616 H2 experienced different thermal-mechanical treatments before thermal aging experiments. NF616 H1 was normalized and tempered, and NF616 H2 was normalized and tempered and cold-rolled. By examining these two heats, we evaluated the effects of thermal-mechanical treatments on material microstructures and associated mechanical properties during long-term aging at elevated temperatures. Thermal aging experiments at different temperatures and periods of time have been completed: 550 C for up to 5000 h, 600 C for up to 7500 h, and 650 C for more than 10,000 h. Tensile properties were measured on thermally aged specimens and aging effect on tensile behavior was assessed. Effects of thermal aging on deformation and failure mechanisms were investigated by using in-situ straining technique with simultaneous synchrotron XRD measurements

Environmentally Assisted Cracking in Light Water Reactors Annual Report January - December 2005.A

This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data, obtained in the pressurized water reactor environment, are presented on Ni-alloy welds prepared in the laboratory or obtained from the nozzle-to-pipe weld of the V. C. Summer reactor. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of these materials to environmentally enhanced cracking under a variety of loading conditions

Environmentally Assisted Cracking in Light Water Reactors - Annual Report, January-December 2001.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level 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 mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several heat treatment conditions

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

Performance of V-Cr-Ti Alloys in a Hydrogen Environment* PERFORMANCE OF V-Cr-Ti ALLOYS IN A HYDROGEN ENVIRONMENT

Abstract A systematic study is underway at Argonne National Laborato~to evaluate the mechanical properties of several V-Cr-Ti alloys after exposure to environments containing hydrogen at various partial pressures. The goal is to correlate the chemistry of the exposure environment with hydrogen uptake by the samples and with the resulting influence on microstructure and tensile properties of the alloys. Other variables examined are specimen cooling rate and synergistic effects, if any, of oxygen and hydrogen on tensile behavior of the alloys. Experiments were conducted to evaluate the effect of pH2 in the range of 3 x 10-6 and-1 torr on tensile properties of two V-Cr-Ti alloys. Up to pH2 of 0.05 torr, negligible effect of H was observed on either maximum engineering stress or uniform and total elongation. However, uniform and total elongation decreased substantially when the alloys were exposed at 500"C t: 1.0 tom of H2 pressure. Preliminary data from sequential exposures of the materials to low-pOZ and several low-pH2 environments did not reveal adverse effects on the maximum engineering stress or on uniform and total elongation when the alloy contained =2000 wppm O and 16 wppm H. Furthermore, tests in H2-exposed specimens, initially annealed at various temperatures, showed that grain-size variation by a factor of =2 had little or no effect on tensile properties. Also, specimen cooling rate had a small effect, if any, on the tensile properties of the alloy

UNIAXIAL CREEP BEHAVIOR OF V-4Cr-4Ti ALLOY* UNIAXIAL CREEP BEHAVIOR OF V-4Cr-4Ti ALLOY UNIAXIAL CREEP BEHAVIOR OF V-4Cr-4Ti ALLOY

Abstract We are undertaking a systematic study at Argonne National Laboratory to evaluate the uniaxial creep behavior of V-Cr-Ti alloys in a vacuum environment as a function of temperature in the range of 650-800°C and at applied stress levels of 75-380 MPa. Creep strain in the specimens is measured by a linear-variable-differential transducer, which is attached between the fixed and movable pull rods of the creep assembly. Strain is measured at sufficiently frequent intervals during testing to define the creep strain/time curve. A linear least-squares analysis function is used to ensure consistent extraction of minimum creep rate, onset of tertiary creep, and creep strain at the onset of tertiary creep. Creep test data, obtained at 650, 700, 725, and 800°C, showed power-law creep behavior. Extensive analysis of the tested specimens is conducted to establish hardness profiles, oxygen content, and microstructural characteristics. The data are also quantified by the Larson-Miller approach, and correlations are developed to relate time to rupture, onset of tertiary creep, times for 1 and 2% strain, exposure temperature, and applied stress

Performance of V-Cr-Ti alloys in a hydrogen environment

A systematic study is underway at Argonne National Laboratory to evaluate the mechanical properties of several V-Cr-Ti alloys after exposure to environments containing hydrogen at various partial pressures. The goal is to correlate the chemistry of the exposure environment with hydrogen uptake by the samples and with the resulting influence on microstructures and tensile properties of the alloys. Other variables examined are specimen cooling rate and synergistic effects, if any, of oxygen and hydrogen on tensile behavior of the alloys. Experiments were conducted to evaluate the effect of pH{sub 2} in the range of 3 x 10{sup {minus}6} and 1 torr on tensile properties of two V-Cr-Ti alloys. Up to pH{sub 2} of 0.05 torr, negligible effect of H was observed on either maximum engineering stress or uniform and total elongation. However, uniform and total elongation decreased substantially when the alloys were exposed at 500 C to 1.0 torr of H{sub 2} pressure. Preliminary data from sequential exposures of the materials to low-pO{sub 2} and several low-pH{sub 2} environments did not reveal adverse effects on the maximum engineering stress or on uniform and total elongation when the alloy contained {approx} 2,000 wppm O and 16 wppm H. Furthermore, tests in H{sub 2}-exposed specimens, initially annealed at various temperatures, showed that grain-size variation by a factor of {approx} 2 had little or no effect on tensile properties. Also, specimen cooling rate had a small effect, if any, on the tensile properties of the alloy