Stress Corrosion Cracking of Type 422 Stainless Steel

This paper presents the results of SCC tests of quenched and tempered martensitic Type 422 SS in neutral and acidic aqueous environments at ambient temperature and 90oC.The susceptibility of smooth and notched tensile specimens to SCC was evaluated by using constant load (CL) and slow strain rate (SSR) test methods. During CL testing, a calibrated proof ring was used to apply a constant load to the test specimen. The magnitude of the applied stress was based on ambient temperature yield strength of the material. On the contrary, the test specimen during SSR testing was continuously strained in tension until fracture at a strain rate of 3.3*10-6sec-1.The fractographic evaluations of all broken specimens were performed by using scanning electron microscopy (SEM)

Environment-Induced Degradations in a Target Structural Material for Transmutation Applications

This investigation is focused on the evaluation of stress corrosion cracking (SCC) and localized corrosion behavior of Type 422 stainless steel in aqueous environments at ambient and elevated temperature. The results of constant load SCC testing using smooth specimens showed no failure in the neutral solution but a threshold stress of 97 ksi was observed in the 90°C acidic environment. SCC testing by the slow-strain-rate test method indicate that the time-to-failure, true failure stress and ductility parameters were gradually reduced with increasing temperature, showing more pronounced effect in the acidic solution. The application of a controlled cathodic potential showed further reduction in the magnitude of these parameters indicating a detrimental effect on the cracking due to hydrogen generation. The results of cyclic potentiodynamic polarization testing revealed pits and crevices on the specimen surface, showing more active (negative) critical pitting potential in the 90°C acidic solution, as expected. Metallographic and fractographic evaluations showed secondary cracks at the gage section and a combination of ductile/brittle failures at the primary fracture face, respectively

Stress Corrosion Cracking of Type 422 Stainless Steel for Applications in Spallation-Neutron-Target Systems

Introduction • This research program is aimed at evaluating different types of environment-induced degradation of candidate target materials for applications in transmutation of spent nuclear fuels (SNF). • Transmutation refers to the elimination of long-lived actinides and fission products from SNF. Objectives • Evaluate susceptibility of candidate target materials to stress corrosion cracking (SCC) and localized corrosion (pitting and crevice) in neutral and acidic aqueous environments at ambient and elevated temperatures • Determine the extent and morphology of cracking in tested materials as functions of experimental and environmental variables including pH, temperature, loading conditions and specimen geometry • Develop mechanistic understanding of degradations based on the experimental dat

Effects of Environmental Variables and Stress Concentration on Cracking of Spallation Target Materials

This paper presents the results of stress corrosion cracking (SCC) studies of two martensitic target materials, namely Alloy EP-823 and Type 422 stainless steel. The susceptibility to SCC was evaluated by using constantload and slow-strain-rate (SSR) test techniques in neutral (pH: 6-7) and acidic (pH: 2-3) aqueous solutions at ambient temperature and 90oC. A proof ring was used to apply tensile load to the smooth cylindrical specimen for 30 days in constant-load testing. For SSR testing, the specimen was strained in tension until fracture at a strain rate of 3.3 x 10-6 sec-1

Thermomechanical Properties of Neutron Irradiated Al\u3csub\u3e3\u3c/sub\u3eHf-Al Thermal Neutron Absorber Materials

A thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4Tm. Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al3Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al3Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al3Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of −6.3% for the annealed irradiated 36.5 vol% specimen

Gas-Fast Reactor Fuel Fabrication

The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum

Small-Scale Specimen Testing of Monolithic U-Mo Fuel Foils

The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength

Ion Irradiation and Examination of Additive Friction Stir Deposited 316 Stainless Steel

This study explored solid-state additive friction stir deposition (AFSD) as a modular manufacturing technology, with the aim of enabling a more rapid and streamlined on-site fabrication process for large meter-scale nuclear structural components with fully dense parts. Austenitic 316 stainless steel (SS) is an excellent candidate to demonstrate AFSD, as it is a commonly-used structural material for nuclear applications. The microstructural evolution and concomitant changes in mechanical properties after 5 MeV Fe++ ion irradiation were studied comprehensively via transmission electron microscopy and nanoindentation. AFSD-processed 316 SS led to a fine-grained and ultrafine-grained microstructure that resulted in a simultaneous increase in strength, ductility, toughness, irradiation resistance, and corrosion resistance. The AFSD samples did not exhibit voids even at 100 dpa dose at 600 °C. The enhanced radiation tolerance as compared to conventional SS was reasoned to be due to the high density of grain boundaries that act as irradiation-induced defect sinks

Degradations of Type 422 Stainless Steel in Aqueous Environments

The susceptibility of Type 422 stainless steel (UNS S42200) to stress corrosion cracking (SCC) and localized corrosion was determined in neutral and acidic aqueous solutions at ambient and elevated temperatures. No failures were observed in the neutral solution at constant load. SCC testing by the slow-strain-rate technique revealed reduced ductility, time-to-failure and true failure stress due to the combined effect of elevated temperature and lower pH. These parameters were further reduced due to the cathodic charging. The localized corrosion studies using the cyclic potentiodynamic polarization technique showed pits and crevices in all specimens. Metallographic and fractographic evaluations showed secondary cracks along the gage section, and a combination of ductile and brittle failures at the primary fracture face of the tested specimen, respectively, depending upon the test environment