Showing papers by "Gary S. Was published in 2011"

Towards an integrated materials characterization toolbox

Abstract: The material characterization toolbox has recently experienced a number of parallel revolutionary advances, foreshadowing a time in the near future when material scientists can quantify material structure evolution across spatial and temporal space simultaneously. This will provide insight to reaction dynamics in four-dimensions, spanning multiple orders of magnitude in both temporal and spatial space. This study presents the authors’ viewpoint on the material characterization field, reviewing its recent past, evaluating its present capabilities, and proposing directions for its future development. Electron microscopy; atom probe tomography; x-ray, neutron and electron tomography; serial sectioning tomography; and diffraction-based analysis methods are reviewed, and opportunities for their future development are highlighted. Advances in surface probe microscopy have been reviewed recently and, therefore, are not included [D.A. Bonnell et al.: Rev. Modern Phys. in Review]. In this study particular attention is paid to studies that have pioneered the synergetic use of multiple techniques to provide complementary views of a single structure or process; several of these studies represent the state-of-the-art in characterization and suggest a trajectory for the continued development of the field. Based on this review, a set of grand challenges for characterization science is identified, including suggestions for instrumentation advances, scientific problems in microstructure analysis, and complex structure evolution problems involving material damage. The future of microstructural characterization is proposed to be one not only where individual techniques are pushed to their limits, but where the community devises strategies of technique synergy to address complex multiscale problems in materials science and engineering.

Irradiation-assisted stress corrosion cracking

Abstract: Irradiation-assisted stress corrosion cracking (IASCC) is aptly named since the effect of irradiation is to enhance an inherent susceptibility to stress corrosion cracking (SCC). This chapter introduces the basic SCC dependencies in austenitic stainless steels and nickel alloys under unirradiated conditions and then describes how they are accentuated or diminished as a result of radiation. SCC results from a confluence of stress, microstructure, and water chemistry, and each is affected by irradiation. With increasing plant operation and improved laboratory capability, it has been concluded that true immunity to SCC growth apparently does not exist in common engineering materials although different conditions can produce large changes in SCC susceptibility. As nuclear power plants operate longer, an increased incidence of SCC can be expected unless active mitigation steps are taken.

Relationship between localized strain and irradiation assisted stress corrosion cracking in an austenitic alloy

Abstract: Irradiation assisted stress corrosion cracking may be linked to the local slip behavior near grain boundaries that exhibit high susceptibility to cracking. Fe–13Cr–15Ni austenitic steel was irradiated with 2 MeV protons at 360 °C to 5 dpa and strained in 288 °C simulated BWR conditions. Clusters of grains from the experiment were created in an atomistic simulation and then virtually strained using molecular dynamic simulation techniques. Cracking and grain orientation data were characterized in both the experiment and the simulation. Random high angle boundaries with high surface trace angles with respect to the tensile direction were found to be the most susceptible to cracking. Grain boundary cracking susceptibility was also found to correlate strongly with slip continuity, indicating that the strain accommodation at the boundary is related to cracking resistance. Higher cracking susceptibility was also found at grain boundaries adjacent to grains with low Schmid factor or high Taylor factor. The basic trends reported here are supported by both the experiments and the simulations.

High-Temperature Oxidation of Alloy 617 in Helium Containing Part-Per-Million Levels of CO and CO2 as Impurities

Abstract: The objective of this study was to determine the mechanisms of carburization and decarburization of alloy 617 in impure helium. To avoid the coupling of multiple gas/metal reactions that occurs in impure helium, oxidation studies were conducted in binary He + CO + CO2 gas mixtures with CO/CO2 ratios of 9 and 1272 in the temperature range 1123 K to 1273 K (850 °C to 1000 °C). The mechanisms were corroborated through measurements of oxidation kinetics, gas-phase analysis, and surface/bulk microstructure examination. A critical temperature corresponding to the equilibrium of the reaction 27Cr + 6CO ↔ 2Cr2O3 + Cr23C6 was identified to lie between 1173 K and 1223 K (900 °C and 950 °C) at CO/CO2 ratio 9, above which decarburization of the alloy occurred via a kinetic competition between two simultaneous surface reactions: chromia formation and chromia reduction. The reduction rate exceeded the formation rate, preventing the growth of a stable chromia film until carbon in the sample was depleted. Surface and bulk carburization of the samples occurred for a CO/CO2 ratio of 1272 at all temperatures. The surface carbide, Cr7C3, was metastable and nucleated due to preferential adsorption of carbon on the chromia surface. The Cr7C3 precipitates grew at the gas/scale interface via outward diffusion of Cr cations through the chromia scale until the activity of Cr at the reaction site fell below a critical value. The decrease in activity of chromium triggered a reaction between chromia and carbide: Cr2O3 + Cr7C3 → 9Cr+3CO, which resulted in a porous surface scale. The results show that the industrial application of the alloy 617 at T > 1173 K (900 °C) in impure helium will be limited by oxidation.

Oxidation of a Proton-Irradiated 316 Stainless Steel in Simulated BWR NWC Environment

Abstract: A proton-irradiated SUS316 stainless steel was exposed to the simulated BWR NWC environment for 70 hours during a constant extension rate tensile test and the resulted oxide film was examined using transmission electron microscopy The oxide film on both the unirradiated and irradiated parts of the sample consists of an outer layer of hematite particles and an inner layer of (Fe, Cr, Ni)3O4 spinel Formation of hematite under BWR NWC condition is consistent with the predication by the potential-pH diagram Both the outer layer and the inner layer of the oxide film show a strong dependence on grain orientation Some grains exhibit an inner layer thickness of 40–100 nm while some others have barely any oxidation Persistent damage induced by proton irradiation did not show a strong influence on the oxidation process as the thickness structure and compositions of the oxide film on both the unirradiated and irradiated parts of the sample were very similar

Chromia-Assisted Decarburization of W-Rich Ni-Based Alloys in Impure Helium at 1273 K (1000 °C)

Abstract: The microstructure and surface stability of two experimental W-rich Ni-based alloys have been studied at 1273 K (1000 °C) in an impure-He environment containing only CO and CO2 as impurities. The alloy Ni-2.3Al-12Cr-12W contained 0.08 wt pct carbon in solution, whereas the second alloy Ni-2.3Al-3Mo-12Cr-12Co-12W contained M6C carbides at the same carbon level. Both alloys, which were preoxidized with ~2.3 μm Cr2O3 layer, were decarburized completely within 50 hours of exposure to the helium gas mixture at 1273 K (1000 °C) via the following chromia-assisted decarburization reaction: Cr2O3 (s) + 3Calloy (s) → 2Cr (s) + 3CO (g). Microstructural observations, bulk carbon analysis, and microprobe measurements confirmed that the carbon in solid solution reacted with the surface chromium oxide resulting in the simultaneous loss of chromia and carbon. The Cr produced by the decomposition of the Cr2O3 diffused back into the alloy, whereas CO gas was released and detected by a gas chromatograph. Once the alloy carbon content was reduced to negligible levels, subsequent exposure led to the uninterrupted growth of Cr2O3 layer in both alloys. In the preoxidized alloys, chromia-assisted decarburization rates were slower for an alloy containing carbides compared with the alloy with carbon in solid solution only. The formation of Cr2O3 is shown to be the rate-limiting step in the chromia-assisted decarburization reaction. Exposure of as-fabricated alloys to the impure-He environment led to the formation of a thin layer of Al2O3 (<1 μm) between the substrate and surface Cr2O3 oxide that inhibited this decarburization process by acting as a diffusion barrier.

Role of Slip Behavior in the Irradiation Assisted Stress Corrosion Cracking in Austenitic Steels

Abstract: Irradiation assisted stress corrosion cracking appears to be linked to the localization of slip into dislocation channels. Three austenitic steels with varying degrees of cracking susceptibility were irradiated with 2 MeV protons at 360°C to 5 dpa and strained in 288°C simulated BWR conditions. Deformation behavior was characterized by Schmid factors, resolved shear stresses, slip continuity across grain boundaries, and the angle between dislocation channels and the cracked boundaries. Cracking susceptibility was found to correlate with the dislocation channel properties, such as the resolved shear stress and slip continuity at grain boundaries. Higher cracking susceptibility was found at grain boundaries perpendicular to the tensile axis and adjacent to low Schmid factor grains, which have high normal stresses acting on the boundary. Localized deformation and high normal stress have significant roles in IASCC, though they do not fully describe cracking susceptibility.

The Key Factors Affecting Crack Growth Behavior of Neutron-Irradiated Austenitic Alloys

Abstract: The key factors affecting crack growth behavior of neutron irradiated stainless steel were investigated in this study. A crack growth rate (CGR) test was conducted under accurate constant K control on a neutron-irradiated 8 mm RCT specimen (9.6 dpa) of high purity 316L stainless steel with Hf addition in simulated BWR (NWC, HWC) and PWR environments at 320°C and 288°C.The effects of water chemistry, electrochemical corrosion potential (ECP), stress intensity factor (K), and temperature on CGR were examined. In addition, the CGR results from this test were integrated with those reported previously in the Cooperative IASCC Research (CIR) program to compared data and determine the effect of the addition of Hf on crack growth rate.

Conducting Well-Controlled Ion Irradiations To Understand Neutron Irradiation Effects In Materials

Abstract: A firm understanding of the effect of radiation on materials is required to develop predictive models of materials behavior in‐reactor and provide a foundation for creating new, more radiation‐tolerant materials. Ion irradiation can serve this purpose for nuclear reactor components and is becoming a key element of materials development for advanced nuclear reactors. Ion irradiations can be conducted quickly, at low cost, and with precise control over irradiation temperature, temperature uniformity, dose rate, dose uniformity and total dose. During proton irradiations the 2σ (twice the standard deviation) of the sample temperature is generally below ∼7 °C, the dose rate variation ∼3%, the dose uncertainty ∼3%, and there is an excellent temperature and dose uniformity across the irradiated area. In this article, we describe the experimental setup and irradiation procedure used to conduct well‐controlled ion irradiations at the University of Michigan.A firm understanding of the effect of radiation on materials is required to develop predictive models of materials behavior in‐reactor and provide a foundation for creating new, more radiation‐tolerant materials. Ion irradiation can serve this purpose for nuclear reactor components and is becoming a key element of materials development for advanced nuclear reactors. Ion irradiations can be conducted quickly, at low cost, and with precise control over irradiation temperature, temperature uniformity, dose rate, dose uniformity and total dose. During proton irradiations the 2σ (twice the standard deviation) of the sample temperature is generally below ∼7 °C, the dose rate variation ∼3%, the dose uncertainty ∼3%, and there is an excellent temperature and dose uniformity across the irradiated area. In this article, we describe the experimental setup and irradiation procedure used to conduct well‐controlled ion irradiations at the University of Michigan.

Stress Corrosion Crack Initiation Susceptibility of Irradiated Austenitic Stainless Steels

Abstract: The susceptibility of neutron irradiated austenitic stainless steels to the initiation of irradiation-assisted stress corrosion cracking (IASCC) was assessed. Solution annealed (SA), high purity (HP) type 304 stainless steel with and without additions of Mo and Si, and HP type 316L +Hf were strained by constant extension rate testing (CERT) in simulated 288°C BWR NWC at a rate of 3.5 × 10−7/s. CERT test data and fracture analysis showed that IASCC susceptibility increased in order of HP304, HP304+Mo, HP316L+Hf, and HP304+Si. This trend was also observed when comparing fracture surfaces of the same alloys tested by CERT in BWR NWC after proton irradiation. Differences were insignificant among reported crack growth rate (CGR) values for the same neutron irradiated alloys, and no connection between crack initiation and CGR was confirmed from the alloys tested.

Grain Boundary Oxidation and Stress Corrosion Cracking in Nickel‐Base Alloys Strained in Supercritical Water

Abstract: The objective of this study is to link grain boundary oxidation to grain boundary cracking in nickel-base alloys using the stability of Ni and NiO as a frame of reference. Accelerated stress corrosion cracking tests and exposures were conducted on alloy 600, Ni-9Fe, and Ni-9Fe-5Cr (LCr) in constant extension rate mode in supercritical water (SCW) at 400°C using dissolved hydrogen concentrations of 47 cc/kg and 200 cc/kg to control the stability of NiO and Ni respectively. Unstressed samples of Ni-9Fe exposed in the NiO stable regime and LCr exposed in both the Ni and NiO stable regimes show grain boundary oxides extending several microns below the sample surface. Constant extension rate tensile test results showed that cracking was more pronounced in samples where intergranular oxides were able to form, except in alloy 600 where no intergranular oxides formed. Comparison with oxide penetration from 400°C hydrogenated steam revealed that the supercritical water environment was more aggressive, but does not suggest a different mechanism of cracking is operating.

Effect of Environment and Prestrain on IASCC of Austenitic Stainless Steels

Abstract: The effect of environment and prestrain on IASCC of austenitic stainless steels was investigated in post-irradiation CERT tests in Ar at 288°C and simulated BWR NWC. Two alloys susceptible to IASCC were selected for this study, which are commercial alloy 304 and high purity Fe-15Crl2Ni. Samples were irradiated to 5 dpa with 2 MeV protons at 360°C and strained to about 2% in Ar at 288°C. The results showed that the water environment is the key to inducing IASCC at 288°C. No intergranular cracking was observed in either alloy following straining in Ar up to ~3.5%. However, cracking occurred once the samples were subsequently strained to an additional 1% in simulated BWR NWC. Cracks tended to be long and extend over many grain boundaries rather than single grain facets, perhaps due to the prestrain in Ar. Cracking induced by 0.2% plastic strain in simulated BWR NWC in CP304 with 3.4% prestrain in Ar was also observed. Normal stress is critical in determining the crack initiation location in CP304 and Fe-15Cr12Ni when strained in simulated BWR NWC because the cracked grain boundaries were preferentially aligned perpendicular to the tensile direction.