Strain rate dependence of impact properties of sintered 316L stainless steel

This paper uses a material testing system (MTS) and a compressive split-Hopkinson bar to investigate the impact behaviour of sintered 316L stainless steel at strain rates ranging from 10⁻³ s⁻¹ to 7.5 × 10³ s⁻¹. It is found that the true stress, the rate of work hardening and the strain rate sensitivity vary significantly as the strain rate increases. The flow behaviour of the sintered 316L stainless steel can be accurately predicted using a constitutive law based on Gurson’s yield criterion and the flow rule proposed by Khan, Huang and Liang (KHL). Microstructural observations reveal that the degree of localized grain deformation increases, but the pore density and the grain size decrease, with increasing strain rate. Adiabatic shear bands associated with cracking are developed at strain rates higher than 5.6 × 10³ s⁻¹. The fracture surfaces exhibit ductile dimples. The depth and density of these dimples decrease with increasing strain rate.     

Study of low-cycle fatigue of AL6XN austenitic stainless steel

The low-cycle fatigue (LCF) behaviors of the AL6XN austenitic stainless steel were investigated at strain rates of 3.3 × 10⁻⁵ to 3.3 × 10⁻³ s⁻¹ and at different temperatures, respectively. The weakened cyclic softening, negative strain-rate-stress response and anomalous temperature-dependence of cycle stress reflect the dynamic strain aging (DSA) hardening during LCF at elevated temperatures. Electron microscopy observations revealed that the dislocation structure changes from the cellular structure at room temperature to the planar slip band, serving as crack initiation sites in the regime of DSA. The DSA hardening results in the reduction of fatigue resistance at elevated temperature via reducing the crack initiation and propagation life.     

A study of hot deformation behavior and microstructural characterization of Mo-TZM alloy

Hot compression testing of Mo-TZM (Mo-0.5Ti-0.1Zr-0.02C) alloy was carried out between 600 and 900 °C employing strain rates from 0.001 s⁻¹ to 1 s⁻¹. Both the constant strain rate and strain rate change test results showed that Mo-TZM possesses low strain rate sensitivity in this temperature range. Activation energy calculated by using strain rate change data and plotting temperature compensated strain rate (Z) vs. shear modulus corrected flow stress (σ/G) was found to be 290 kJ/mol. Electron back scattered diffraction (EBSD) and transmission electron microscopic (TEM) results obtained from the rapidly cooled deformed specimens revealed the formation of subgrains. Flow stress-plastic strain results and misorientation angles between subgrains showed an anomalous behavior at 800 °C.     

Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

Proton irradiation effect on microstructure, strain localization and iodine-induced stress corrosion cracking in Zircaloy-4

The radiation-induced microstructure, strain localization, and iodine-induced stress corrosion cracking (I-SCC) behaviour of recrystallized Zircaloy-4 proton-irradiated to 2 dpa at 305 °C was examined. <a> type dislocation loops having 1/3〈1 1  0〉 Burgers vector and a mean diameter and density of, respectively, 10 nm and 17 × 10²¹ m⁻³ were observed while no Zr(Fe,Cr)₂ precipitates amorphization or Fe redistribution were detected after irradiation. After transverse tensile testing to 0.5% macroscopic plastic strain at room temperature, almost exclusively basal channels were imaged. Statistical Schmid factor analysis shows that irradiation leads to a change in slip system activation from prismatic to basal due to a higher increase of critical resolved shear stresses for prismatic slip systems than for basal slip system. Finite element calculations suggest that dislocation channeling occurs in the irradiated proton layer at an equivalent stress close to 70% of the yield stress of the irradiated material, i.e. while the irradiated layer is still in the elastic regime for a 0.5% applied macroscopic plastic strain. Comparative constant elongation rate tensile tests performed at a strain rate of 10⁻⁵ s⁻¹ in iodized methanol solutions at room temperature on specimens both unirradiated and proton-irradiated to 2 dpa demonstrated a detrimental effect of irradiation on I-SCC.     

Characterization of hot deformation behavior of Zr-1Nb-1Sn alloy

The hot deformation behavior of β-quenched Zr-1Nb-1Sn was studied in the temperature range 650-1050 °C and strain rate range 0.001-100 s⁻¹ using processing maps. These maps revealed three different domains: a domain of dynamic recovery at temperatures <700 °C and at strain rates <3 × 10⁻³ s⁻¹, a domain of dynamic recrystallization in the temperature range 750-950 °C and at strain rates <10⁻² s⁻¹ with a peak at 910 °C and 10⁻³ s⁻¹ (in α + β phase field), and a domain of large-grain superplasticity in the β phase field at strain rates <10⁻² s⁻¹. In order to identify the rate controlling mechanisms involved in these domains, kinetic analysis was carried out to determine the various activation parameters. In addition, the processing maps showed a regime of flow instability spanning both α + β and β phase fields. The hot deformation behavior of Zr-1Nb-1Sn was compared with that of Zr, Zr-2.5Nb and Zircaloy-2 to bring out the effects of alloy additions.     

Analysis of tensile deformation and failure in austenitic stainless steels: Part I - Temperature dependence

This paper describes the temperature dependence of deformation and failure behaviors in the austenitic stainless steels (annealed 304, 316, 316LN, and 20% cold-worked 316LN) in terms of equivalent true stress-true strain curves. The true stress-true strain curves up to the final fracture were calculated from tensile test data obtained at −150 to 450 °C using an iterative finite element method. Analysis was largely focused on the necking and fracture: key parameters such as the strain hardening rate, equivalent fracture stress, fracture strain, and tensile fracture energy were evaluated, and their temperature dependencies were investigated. It was shown that a significantly high strain hardening rate was retained during unstable deformation although overall strain hardening rate beyond the onset of necking was lower than that of the uniform deformation. The fracture stress and energy decreased with temperature up to 200 °C and were nearly saturated as the temperature came close to the maximum test temperature 450 °C. The fracture strain had a maximum at −50 to 20 °C before decreasing with temperature. It was explained that these temperature dependencies of fracture properties were associated with a change in the dominant strain hardening mechanism with test temperature. Also, it was seen that the pre-straining of material has little effect on the strain hardening rate during necking deformation and on fracture properties.     

Effects of stress on radiation hardening and microstructural evolution in A533B steel

Bent specimens of A533B steel (0.16 wt% Cu) were irradiated at 290 °C to 1 dpa with 6.4 MeV Fe³⁺ ions. Calculated tensile stresses at the irradiated surface were set to 0, 250, 500 and 750 MPa. The specimens were subjected to hardness measurements, transmission electron microscopy (TEM) observations and three-dimensional atom probe (3DAP) analysis. The radiation-induced hardening decreased with increasing stress to 500 MPa which was near the yield strength. TEM and 3DAP results showed that well-defined dislocation loops and solute clusters were formed. The diameter of dislocation loops increased and the number density decreased when the stress was applied, whereas the diameter and number density of solute clusters decreased. The hardening was mainly attributed to solute cluster formation. Application of tensile stress would control hardening by suppressing the solute cluster nucleation and growth.     

Serrated flow behavior in AL6XN austenitic stainless steel

Serrated flow behavior of the AL6XN austenitic stainless steel has been investigated at different temperatures and strain rates. The results show the serrated flow, peak/plateau in flow stress and negative strain rate sensitivity appearing in tensile deformation of the AL6XN steel at 773-973 K and 3.3 × 10⁻⁵-3.3 × 10⁻³ s⁻¹ (excluding 873 K, 3.3 × 10⁻⁵ s⁻¹), suggesting the occurrence of dynamic strain aging (DSA). The activation energy for type-A and -(A + B) serrations was calculated to be 304 kJ/mol and diffusion of substitutional solutes, such as chromium and molybdenum is considered as the mechanism of serrated flow. TEM observations further revealed a typical planar slip mode in the regime of DSA of the deformed AL6XN steel.     

Study of silver diffusion in silicon carbide

Diffusion of silver in 6H-SiC and polycrystalline CVD-SiC was investigated using α-particle channeling spectroscopy and electron microscopy. Fluences of 2 × 10¹⁶ cm⁻² of ¹⁰⁹Ag⁺ were implanted with an energy of 360 keV at room temperature, at 350 °C and 600 °C, producing an atomic density of approximately 2% at the projected range of about 110 nm. The broadening of the implantation profile and the loss of silver through the front surface during vacuum annealing at temperatures up to 1600 °C was determined. Fairly strong silver diffusion was observed after an initial 10 h annealing period at 1300 °C in both polycrystalline and single crystalline SiC, which is mainly due to implant induced radiation damage. After further annealing at this temperature no additional diffusion took place in the 6H-SiC samples, while it was considerably reduced in the CVD-SiC. The latter was obviously due to grain boundary diffusion and could be described by the Fick diffusion equation. Isochronal annealing of CVD-SiC up to 1400 °C exhibited an Arrhenius type temperature dependence, from which a frequency factor D_o ∼ 4 × 10⁻¹² m² s⁻¹ and an activation energy E_a ∼ 4 × 10⁻¹⁹ J could be extracted. Annealing of 6H-SiC above 1400 °C shifted the silver profile without any broadening towards the surface, where most of the silver was released at 1600 °C. Electron microscopy revealed that this process was accompanied by significant re-structuring of the surface region. An upper limit of D < 10⁻²¹ m² s⁻¹ was estimated for 6H-SiC at 1300 °C.     

Liquid metal embrittlement susceptibility of ferritic-martensitic steel in liquid lead alloys

The susceptibility of the ferritic-martensitic steels T91 and EUROFER97 to liquid metal embrittlement (LME) in lead alloys has been examined under various conditions. T91, which is currently the most promising candidate material for the high temperature components of the future accelerator driven system (ADS) was tested in liquid lead bismuth eutectic (LBE), whereas the reduced activation steel, EUROFER97 which is under consideration to be the structural steel for fusion reactors was tested in liquid lead lithium eutectic. These steels, similar in microstructure and mechanical properties in the unirradiated condition were tested for their susceptibility to LME as function of temperature (150-450 °C) and strain rate (1 × 10⁻³-1 × 10⁻⁶ s⁻¹). Also, the influence of pre-exposure and surface stress concentrators was evaluated for both steels in, respectively, liquid PbBi and PbLi environment. To assess the LME effect, results of the tests in liquid metal environment are compared with tests in air or inert gas environment. Although both unirradiated and irradiated smooth ferritic-martensitic steels do not show any or little deterioration of mechanical properties in liquid lead alloy environment compared to their mechanical properties in gas as function of temperature and strain rate, pre-exposure or the presence of surface stress concentrators does lead to a significant decrease in total elongation for certain test conditions depending on the type of liquid metal environment. The results are discussed in terms of wetting enhanced by liquid metal corrosion or crack initiation processes.     

Thermal and structural properties of low-fluence irradiated graphite

The release of Wigner energy from graphite irradiated by fast neutrons at a TRIGA Mark II research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry / synchrotron powder X-ray diffraction between 25 and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from 5.67 × 10²⁰ to 1.13 × 10²² n m⁻² at a fast-neutron flux (E > 0.1 MeV) of 7.88 × 10¹⁶ n m⁻² s⁻¹ and at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 1.2 to 21.8 J g⁻¹ and a Wigner energy accumulation rate of 1.9 × 10⁻²¹ J g⁻¹ n⁻¹ m². The differential-scanning-calorimeter curves exhibit, in addition to the well known peak at ∼200 °C, a pronounced fine structure consisting of additional peaks at ∼150, ∼230, and ∼280 °C. These peaks correspond to activation energies of 1.31, 1.47, 1.57, and 1.72 eV, respectively. Crystal structure of the samples is intact. The dependence of the c lattice parameter on temperature between 25 and 725 °C as determined by Rietveld refinement leads to the expected microscopic thermal expansion coefficient along the c axis of ∼26 × 10⁻⁶ °C⁻¹. At 200 °C, coinciding with the maximum in the differential-scanning-calorimeter curves, no measurable changes in the rate of thermal expansion have been detected - unlike its decrease previously seen in more highly irradiated graphite.     

Study of iodine diffusion in silicon carbide

Diffusion of iodine in 6H-SiC and polycrystalline CVD-SiC was investigated using Rutherford backscattering spectroscopy and electron microscopy. A fluence of 1 × 10¹⁶ cm⁻² of ¹²⁷I⁺ was implanted with an energy of 360 keV at room temperature, producing an amorphous surface layer of approximately 220 nm thickness. The implantation profile reached an atomic density of approximately 1.3% at the projected range of about 95 nm. Broadening of the implantation profile and iodine loss through the front surface during isochronal and isothermal vacuum annealing was determined. At a temperature of 1100 °C no iodine loss was observed after 120 h and a diffusion coefficient of less than 10⁻²¹ m² s⁻¹ was extracted from the analysis of profile widths. Relatively strong broadening occurred after 60 h annealing at 1200 °C with the iodine profile extending beyond 300 nm into the bulk, accompanied by a surprisingly modest iodine loss through the surface. Electron microscopic studies reveal a drastic restructuring of the surface region at this temperature, indicating possible chemical reactions between iodine and silicon carbide.     

Permeability of hydrogen and deuterium of Hastelloy XR

Permeation of hydrogen isotope through a high-temperature alloy used as heat exchanger and steam reformer pipes is an important problem in the hydrogen production system connected to be a high-temperature engineering test reactor (HTTR). An experiment of hydrogen (H₂) and deuterium (D₂) permeation was performed to obtain permeability of H₂ and D₂ of Hastelloy XR, which is adopted as heat transfer pipe of an intermediate heat exchanger of the HTTR. Permeability of H₂ and D₂ of Hastelloy XR were obtained as follows. The activation energy E₀ and pre-exponential factor F₀ of the permeability of H₂ were E₀=67.2±1.2 kJ mol⁻¹ and F₀=(1.0±0.2)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5, respectively, in the pipe temperature ranging from 843 K (570 °C) to 1093 K (820 °C). E₀ and F₀ of the permeability of D₂ were respectively E₀=76.6±0.5 kJ mol⁻¹ and F₀=(2.5±0.3)×10⁻⁸ m³(STP) m⁻¹ s⁻¹ Pa^−0.5 in the pipe temperature ranging from 943 K (670 °C) to 1093 K (820 °C).     

Measurement of density distribution of a cracking catalyst in experimental riser with a sampling procedure for gamma ray tomography

By scanning a riser the number of the gamma ray trajectories and the beam width involve temporal, spatial and density resolutions as they are closely correlated parameters. Therefore, evaluation of parameters and their interaction quantification, certainly, are required in the imaging process. Measuring the density distribution of the catalyst from the FCC - fluid cracking catalytic process in an experimental riser in single beam tomographic system, density resolution is evaluated and correlated with spatial resolution. The beam width Δs inside riser is measured and a criterion for determining spatial resolution is proposed. Experiments are carried out to demonstrate resolution effects of three Δs values: 3.30 × 10⁻³, 6.20 × 10⁻³ and 12.00 × 10⁻³ m. The gamma beam profile is modeled and a sampling rate according to Nyquist criterion is analyzed. The 4.3%, 8.1% and 15.6% ratios of Δs/R to internal riser radius are correlated to counting time in the sampling procedure. Results are discussed by comparison with values from literature.     

Nuclear reaction analysis of helium migration in silicon carbide

4H-SiC and 6H-SiC single crystals were implanted at room temperature with 3-MeV ³He ions at a fluence of 1 × 10¹⁶ cm⁻². Analysis of helium migration was carried out with the ³He(d, p)⁴He nuclear reaction. No clear thermally-activated migration in the end-of-range (EOR) region is found below 1100 °C, meaning that helium is strongly trapped probably in helium-vacancy clusters. At 1100 °C and above, a fraction of ³He atoms remains trapped in the clusters, but a significant fraction is detrapped into a broad distribution, which is slightly shifted towards the sample surface. Helium detrapping from the EOR region increases with increasing annealing time and temperature. Moreover, the helium content is not conserved, since a significant fraction of ³He atoms is released out of the sample. Helium out-gassing actually increases with increasing annealing time and temperature, up to about 40% at 1150 °C. No clear difference is found between the 4H-SiC and 6H-SiC polytypes.     

Effect of ion irradiation and indentation depth on the kinetics of deformation during micro-indentation of Zr-2.5%Nb pressure tube material at 25 °C

Micro-indentation creep tests were performed at 25 °C on radial-normal samples cut from Zr-2.5Nb CANDU pressure tube material in both the as-fabricated condition and after irradiation with 8.5 MeV Zr⁺ ions. The average indentation stress, and hence the yield stress, was found to increase with decreasing indentation depth and with increasing levels of ion irradiation. The activation energy of the indentation creep rate and hence the, activation energy of the obstacles that limit the rate of dislocation glide, was independent of indentation depth but increased from ΔG₀ = 0.185 to 0.215 μb³ with increasing ion irradiation damage. The magnitude of the activation energy indicates that ion irradiation introduces a new type of obstacle into the microstructure which reduces the low temperature indentation creep rate of Zr-2.5Nb pressure tubes. This is supported by TEM images showing that Zr⁺ ion irradiation produces small, nanometer size, dislocation loops which act as obstacles to dislocation glide and thus influence both the yield stress and the activation energy of the low-temperature thermal creep of Zr-2.5Nb pressure tube material. These findings suggest that neutron irradiation will have similar effect upon yield stress and low-temperature thermal creep as the Zr⁺ ion irradiation since both create similar crystallographic defects in Zr-2.5Nb pressure tubes.     

An experimental investigation on the quenching of a hot vertical heater by water injection at high flow rate

An experimental investigation on rewetting has been carried out by injecting water from the top of a hot vertical heater. Tests have been performed with varied range of experimental conditions (200-500 °C surface temperatures, constant water flow rates 5.77-30.98 g s⁻¹). Effect of several coolant injection systems on the hydrodynamics of rewetting has been studied. It is observed that for a particular range of flow rate and initial wall temperature (21.58 g s⁻¹, 300 °C) a circumferentially symmetric wet front is observed for the region closer to the coolant injection point even while using sub-cooled water. Rewetting velocity has been calculated from the temperature transients measured during the experiment and was found to vary within 1.0-20.0 cm s⁻¹. Two different rewetting models ( [Sahu et al., 2006] and [Sahu et al., 2008a]) have been used to compare the present experimental data and the comparison is found to be fairly good in both the cases. It has been observed that the flow rate varies linearly with effective Biot number (M) and varies inversely with magnitude of precursory cooling (N) in the present investigation.     

Flux dependent MeV ion induced modification of nano-Ag/Si system

Thin films of Ag (1.5 nm thick) are grown on Si (1 1 1) substrates using evaporation method in high vacuum condition and due to non-wetting nature of silver, isolated islands of mean size ≈12.0 nm have been formed on the surface. Au²⁺ (1.5 MeV) ions have been used to irradiate the above systems at various fluences (5 × 10¹³-1 × 10¹⁵ cm⁻²) at an impact angle of 5° and at a flux of 6.3 × 10¹² cm⁻² s⁻¹ (corresponding to a beam current density of 2.0 μA cm⁻² for Au²⁺ ions). Ion beam induced embedding is observed to begin at a fluence of 1 × 10¹⁴ cm⁻² for this high flux whereas low flux irradiations (current density ≈ 0.02 μA cm⁻²) of Au²⁺ ions under similar irradiation conditions did not yield embedding (impact angle 5°). High resolution transmission electron microscopy measurement showed no mixing in the form of silicide formation. These results are compared with high flux modifications in Au/Si system.     

Characterization of defect accumulation in neutron-irradiated Mo by positron annihilation spectroscopy

Positron annihilation lifetime spectroscopy measurements were performed on neutron-irradiated low carbon arc cast Mo. Irradiation took place in the high flux isotope reactor, Oak Ridge National Laboratory, at a temperature of 80 ± 10 °C. Neutron fluences ranged from 2 × 10²¹ to 8 × 10²⁴ n/m² (E > 0.1 MeV), corresponding to displacement damage levels in the range from 7.2 × 10⁻⁵ to 2.8 × 10⁻¹ displacements per atom (dpa). A high density of submicroscopic cavities was observed in the neutron-irradiated Mo and their size distributions were estimated. Cavities were detected even at a very low-dose of ∼10⁻⁴ dpa. The average size of the cavities did not change significantly with dose, in contrast to neutron-irradiated bcc Fe where cavity sizes increased with increasing dose. It is suggested that the in-cascade vacancy clustering may be significant in neutron-irradiated Mo, as predicted by molecular dynamics simulations.