Radiation damage in vanadium

Transmission electron microscopy has been used to study the microstructure of neutron irradiated single crystal vanadium. These observations revealed the presence of small defect clusters which, upon annealing, grew into resolvable dislocation loops, analyzed as interstitial in nature. The radiation-induced clusters and loops were stabilized during annealing by interaction with impurity precipitates. Damage shells, ascribed to transmutation reactions, were found surrounding certain particles believed to contain boron.     

The oxygen snoek damping and an abnormal peak of neutron-irradiated vanadium

Measurements of low-frequency internal friction and electron microscope observations were made on neutron-irradiated vanadium with various oxygen contents. Irradiation was carried out at about 60°C to a fast fluence of $\text{2 × 10}{}^{17}$ or $\text{5 × 10}{}^{19}$ n/cm² ($\text{E ? 1 MeV}$). The oxygen Snoek damping was decreased by irradiation and post-irradiation annealing below 200 or 250° C, while it began to recover by annealing above this temperature. Complete recovery was attained by 30 min anneal at 450°C in the case of the lower fluence, whereas in the other case it was not observed after the same treatment. The results of electron microscope observations were consistent with those of internal friction measurements. The specimens irradiated to $\text{5 × 10}{}^{19}$ n/cm² showed an abnormal peak after annealing above 250°C near the nitrogen Snoek temperature. The height of this peak, $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, was expressed as $\text{P}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ ∝ exp ($\text{2.72 × 10}{}^3\text{/RT}$) $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$, where $\text{Q}{}^{\mathrm{?1}}{}_{\mathrm{<mtext>max</mtext>}}$ the heiβht of the oxygen Snoek damping after each annealing. The mechanism for radiation-anneal hardening and the abnormal peak were considered in the light of these experiments.     

A saturable-trap analysis of solute segregation applied to oxygen trapping in neutron-irradiated vanadium

The activation energy for the 0.2T_m resistivity annealing stage in neutron-irradiated vanadium containing 61 wt ppm oxygen was determined to be (1.21 ± 0.06) eV. This value is reasonably close to the oxygen diffusion activation energy in vanadium of 1.26–1.28 eV. Thus, an extrinsic mechanism for the 0.2T_m annealing stage is indicated, involving oxygen migration and trapping at radiation-produced defect clusters. A simple saturable trap model for the trapping of interstitial impurity atoms at radiation-produced defect clusters is described. The model is applied to the isothermal annealing curves for vanadium containing oxygen. A good fit to the shape of the annealing curves is obtained and approximate agreement to the measured activation energy is found. However, the model appears to overestimate the amount of oxygen participating in the trapping process.     

Effects of solute elements on hardening and microstructural evolution in neutron-irradiated and thermally-aged reactor pressure vessel model alloys

Nanometer-sized Cu-enriched solute clusters containing Mn, Ni, and Si atoms are considered as the primary embrittling feature in reactor pressure vessel steels. In order to understand the effects of solute atoms Mn, Ni, and Si on hardening and cluster formation, reactor pressure vessel model alloys FeCu, FeCuSi, FeCuNi, and FeCuNiMn were irradiated at 290 °C in a research reactor. Thermal ageing at 450 °C was also carried out to compare with the results in the neutron irradiation. The addition of Mn resulted in larger hardening and higher cluster number density in both thermal ageing and neutron irradiation. In FeCu0.8NiMn alloy, the size distribution of Cu-enriched clusters formed in 62-h thermal ageing (almost peak hardening) was very similar to that formed in the neutron irradiation, indicating they are on a similar growing stage. But the average Ni and Mn composition in clusters formed in neutron irradiation was higher. A good linear relationship between hardening and the square root of cluster volume fraction for both neutron irradiation and thermal ageing data was found.     

Faulted loops in neutron-irradiated zirconium

Faulted loops have been observed in high-purity zirconium irradiated at 723 K to 1.3 × 10²⁵ neutrons/m² (> 0.1 MeV). The transmission electron microscopy characterization of these $\text{1}\text{6}$ 〈202&#x0304;3〉 faulted loops on (0001) is described in detail. It was found that the faulted loops were invariably vacancy in character although the coexisting population of perfect $\text{1}\text{3}$ 〈112&#x0304;0〉 loops was of a mixed interstitial/vacancy nature. The faulted loops were observed in specimens of only two out of five batches of high-purity zirconium irradiated in this experiment. Even in these two specimens, the presence of faulted loops was restricted to the 723 K irradiation temperature; at 673 K only perfect $\text{1}\text{3}$ 〈112&#x0304;0〉 loops were seen.     

Interstitial loops in neutron-irradiated aluminium

A method is presented for the estimation of the solubility of uranium and plutonium in solvent systems composed of two or more low-melting metals. The method presented for the estimation of the activity coefficient of uranium and plutonium in multi-component solvent systems is applicable to the prediction of solubilities and of distribution coefficients between liquid alloys and molten salts. The theoretical basis for the salt transport separation process used in pyrochemical methods for the recovery of irradiated fast breeder reactor fuels is presented. The methods are illustrated by the computation of the solubility of uranium in liquid Zn-Mg and Zn-Mg-Ca alloys and the distribution of uranium between liquid Zn-Mg alloy and molten MgCl₂.     

Radiation synthesis and characteristics of PTFE-g-PSSA ion exchange membrane applied in vanadium redox battery

Radiation-induced grafting of styrene onto polytetrafluoroethylene （PTFE） membranes was studied by a simultaneous irradiation technique. Grafting was carded out using γ-radiation from a ^60Co source at room temperature. Effects of absorbed dose, atmosphere, dose rate, and the concentration of initial monomer on the degree of grafting （DOG） were investigated and the most appropriate grafting condition was obtained. Subsequently, sulphonation of the grafted PTFE membrane （PTFE-g-PS） was carded out and a series of ion exchange membranes （PTFE-g-PSSA） was prepared. Further characterizations of FTIR, TGA, and SEM testified that grafting and sulphonation of the membranes were successfully processed; moreover, grafting of styrene not only occurred in the surface of PTFE membrane, but also in the micropores of the membrane. Ion exchange capacity （IEC） and conductivity were found increase with the grafting yield. The results suggest that at a low dose, such as 17 kGy, the ion exchange membrane （IEM） which will be suitable for vanadium redox battery （VRB） use can be obtained.     

Small-angle grain boundaries in neutron-irradiated Cu-B alloy

Dislocation networks and small-angle grain boundaries were observed to form among damage shells produced around B particles in B-doped Cu specimens neutron irradiated to a thermal dose of $\text{3.0 × 10}{}^{19}\text{n/cm}{}^2$ at about 70°C. The microstructures are formed as a results of the interaction of dislocations generated by strain induced around a damage shell with dislocations generated from the other damage shell. The small-angle grain boundaries are formed to relax the strain induced between closely spaced damage shells.     

Corduroy contrast observations in neutron-irradiated zirconium and zircaloy

Observations of a particular as-irradiated microstructure are reported for Zircaloy and zirconium after neutron fluences of 2 to 25 × 10²⁰ n/cm² at temperatures of 523 to 608 K. Because of the ribbed appearance of the electron microscope images, the term ‘corduroy’ is used to describe the general microstructures encountered. Corduroy is a bulk microstructural feature (of irradiated crystals) with well-defined crystallographic properties. The rows of the corduroy are aligned parallel to traces of basal planes and the structure of corduroy is associated with a displacement vector of the type <110n>, where n ≠ 0. Corduroy is analyzed as being composed of small crystal regions misoriented from each other by small amounts so as to periodically vary the Bragg position of the diffracting planes between two extremes. Corduroy does not represent alignment of the common radiation damage clusters, but may be superposed upon these. Annealing experiments have eliminated the corduroy structure while causing the damage clusters to grow into resolvable loops. Corduroy does not appear to be strongly affected by fluence between 2 and 25 × 10²⁰ n/cm² but is sensitive to irradiation temperature.     

国产VDMOS电离辐射环境中的敏感参数研究

分析了国产VDMOS(Vertical Double-diffusion MOSFET)总剂量试验数据,并比较了辐照环境中不同偏置下电学参数的变化差异,发现新工艺研制的VDMOS器件辐照敏感参数为击穿电压和阈值电压.根据试验结果得到考核VDMOS辐照可靠性的有效试验方法:击穿电压抗辐照考核时应采用漏偏置,阈值电压抗辐照考核时应采用栅偏置.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.