Neutron Irradiated Tungsten Research Articles

Cluster dynamics modeling of accumulation and diffusion of helium in neutron irradiated tungsten

A cluster dynamics model based on rate theory has been developed to study the accumulation and diffusion processes of helium in tungsten under synergistic effects of helium implantation and neutron irradiation. By including self-interstitial atoms, vacancies and helium atoms as well as their clusters and further using more reliable parameters, the evolution of different types of defects with time and depth is investigated. The calculated results are comparable with experiments. The cases of helium plasma corresponding to the first wall and to the divertor are taken into account. The accumulation and diffusion behaviors of helium in tungsten are illustrated by the time and depth dependence of helium concentration in tungsten with or without the neutron irradiation, the contribution of different types of helium clusters/complexes to helium concentration and the depth profiles of different mobile defects and helium–vacancy complexes. It is concluded that the competition of trapping and diffusion effects dominates the behavior of helium atoms in tungsten for these two typical cases. Understanding these mechanisms is important for estimating damages to the plasma-facing materials.

The deuterium depth profile in neutron-irradiated tungsten exposed to plasma

Tungsten samples (99.99% purity from A.L.M.T. Corp., 6 mm in diameter, 0.2 mm in thickness) were irradiated by high-flux neutrons at 50 °C to 0.025 dpa in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Subsequently, the neutron-irradiated tungsten samples were exposed to high-flux deuterium plasmas (ion flux: 1021–1022 m−2 s−1, ion fluence: 1025–1026 m−2) in the Tritium Plasma Experiment at Idaho National Laboratory. This paper reports the results of deuterium depth profiling in neutron-irradiated tungsten exposed to plasmas at 100, 200 and 500 °C via nuclear reaction analysis (NRA). The NRA measurements show that a significant amount of deuterium (>0.1 at.% D/W) remains trapped in the bulk material (up to 5 μm) at 500 °C. Tritium Migration Analysis Program simulation results using the NRA profiles indicate that different trapping mechanisms exist for neutron-irradiated and unirradiated tungsten.

Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

The behavior of D retention for Fe2+-irradiated tungsten with a damage of 0.025–3 dpa was compared with that for neutron-irradiated tungsten with 0.025 dpa. The D2 thermal desorption spectroscopy (TDS) spectra for Fe2+-irradiated tungsten consisted of two desorption stages at 450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an addition desorption stage was found at 750 K. The desorption rate of the major desorption stage at 550 K increased as the displacement damage increased due to Fe2+ irradiation increasing. In addition, the first desorption stage at 450 K was found only for damaged samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced by Fe2+ irradiation, and the first stage should be the accumulation of D in mono-vacancy and the activation energy would be relatively reduced, where the dislocation loop and vacancy is produced. The third one was found only for neutron irradiation, showing the D trapping by a void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at half-maximum of the TDS spectrum. Therefore, it can be said that the D2 TDS spectra for Fe2+-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different from that for the ion-irradiated one.

Microstructure Development in Neutron Irradiated Tungsten Alloys

The aim of this work is to investigate the influence of neutron irradiation condition, especially temperature, on irradiation hardening and microstructure development in irradiated tungsten-rhenium (W-Re) alloys. Neutron irradiations were carried out in JOYO at the range of 400 to 750 � C, up to 1.54 dpa. Micro Vickers hardness tests and micro structural observations using a TEM were performed. Irradiation hardening of WRe alloys irradiated at 538 � C were clearly larger than those irradiated at other temperatures. Fine voids and fine needle- or plate-like precipitates were observed in pure W and W-Re alloys irradiated at 538 � C, respectively, with high number density. The fine complex microstructure seems to be the cause of the characteristic irradiation hardening. [doi:10.2320/matertrans.MBW201025]

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

With the Japan–US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323K to 0.025dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5×1021m−2s−1, ion fluence: 4×1025m−2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Effects of transmutation elements on the microstructural evolution and electrical resistivity of neutron-irradiated tungsten

During fusion reactor operation, transmutation elements such as rhenium (Re) and osmium (Os) are produced from tungsten (W) upon neutron irradiation. Thus, the pure W becomes W–Re or W–Re–Os alloys and its physical properties gradually change. The irradiation hardening, microstructural changes, and physical properties of these transmutation elements of W are here investigated. Tungsten-based model alloys are fabricated and neutron irradiation is performed in the JOYO fast test reactor. The irradiation dose and temperature are 0.17–1.54 dpa and 400–750 °C, respectively. Vickers hardness measurements, microstructural observations, and electrical resistivity measurements are subsequently performed. The effects of the microstructural evolution on the irradiation hardening and electrical resistivity are discussed.

Thermoseparation of Neutron-Irradiated Tungsten from Re and Os

Heterogeneous oxidation has been utilized for one-step conversion of metallic W to WO 3 at ∼750 °C in a stream of air, with simultaneous separation of W from Re and Os impurities which are coproduced during the long neutron irradiation of the W target (∼50 days). The freshly produced WO 3 is readily soluble in a base for use in a 188 W/ 188 Re biomedical generator. The metallic tungsten target, in the form of small pellets (4 x 3 mm diameter x height), was prepared by cold pressing W powder at ∼3 tons/cm 2 . Sintering (vacuum annealing) the pellets at 1200-1400 °C, under 1-10 μTorr pressure provided good mechanical strength. Analysis of the black residue, which is typically encountered in the dissolution of neutron-irradiated WO 3 targets, indicated that 186 W and 187 Re were the main constituents. Further, the specific activity of insoluble 188 W in the residue was ∼200 times smaller than that of soluble 188 W.

Separation of No-Carrier-Added188Re from Neutron-Irradiated Natural Tungsten

Rhenium-188 has been determined to be a suitable therapy radiolable due to its unique half-life and β - emission associated with 155 keV photons and chelation properties. 188 Re is obtained from the decay of parent 188 W in the form of a generator system while 188 W is produced by double neutron capture of 186 W. The yield of generators using alumina for separation is limited due to the low sorption capacity of sorbent. We have exploited precipitation as hydrous tungsten oxide in acid media and redissolution in alkaline media to separate 188 Re from tungsten. The optimized experimental conditions indicate the feasibility of a generator system based on this method

Defect recovery behavior of neutron irradiated molybdenum and tungsten of two purity levels

Abstract The recovery behavior of radiation-induced defects during post-irradiation annealing was studied on molybdenum and tungsten specimens of two different purity levels. An electrical resistivity measurement technique at liquid nitrogen boiling temperature (∼77°K) was used. Irradiation of both materials was conducted in Oak Ridge Research reactor at reactor ambient temperature (∼70°C). The accumulative neutron fluence received was 7.3E+19 neutrons cm−2 (En>l MeV) and 5.1E+20 neutrons cm−2 (thermal). It was found that the number of recovery stages appeared to be independent of either the material or the impurity content. The stages are then believed to be due to the recovery of intrinsic defects and the recovery mechanisms are most likely the same for molybdenum and tungsten on the homologous temperature scale.

Reply to some comments on recovery processes in thermal neutron irradiated tungsten

Reply to some comments on recovery processes in thermal neutron irradiated tungsten

Radiation damage and stage iii defect annealing in thermal neutron irradiated tungsten

The damage and the recovery behavior of defects in thermal neutron irradiated tungsten has been investigated in the temperature range of room temperature to 1200°C, using electrical resistivity measurements and field ion microscope observations. Extensive observations of the isochronal annealing of the irradiated specimens of various purity show that the defects, introduced by irradiation, are completely recovered at around 420°C after stage III recovery for the least pure specimens of ∼ 200 at. ppm impurity, while in the purest specimens of ∼ 1 at.ppm impurity an additional recovery stage (stage IV) is present. Field ion microscope observations show that the motion of interstitials is the dominant process in stage III recovery, whose activation energy, determined by resistivity measurements, varies between 1.68 ± 0.05 and 1.71 ± 0.05 eV, depending upon the purity of the material. The occurrence of stage IV in the purest specimens is explained in terms of the clustering of mobile defects which takes place during stage III recovery.

Recovery in neutron irradiated tungsten

Abstract The recovery of defects in partially and well annealed tungsten has been studied using electrical resistivity technique. The irradiation was carried out at reactor ambient temperature (∼343 K) to a fluence of 1022 n/m2. Isochronal annealing studies of the defects show two stages centred around 0.17Tm and 0.31Tm referred to as Stage III and Stage V respectively. Stage III shows a peak temperature shift to lower temperature for increasing amount of recovery. Isothermal annealing of the defects in Stage III and Stage V has been carried out and activation energy calculated using Meechan-Brinkman analysis. A single value of activation energy (1.69 ± 0.20 eV) and essentially a second order kinetics has been found in Stage III. However, no single value of activation energy could be obtained for Stage V (the activation energy varies from 2.5 to 3.6 eV). On the basis of the activation energy, reaction order and the negative shift in the peak temperature, Stage III has been attributed to arise due to migra...

Tensile properties of neutron irradiated TZM and tungsten

Abstract The effect of neutron irradiation on the elevated temperature tensile properties of TZM and tungsten has been experimentally determined. Specimens were irradiated at a temperature of approximately 385°C to fluences of 0.4 and 0.9 × 1022n/cm2 (E > 0.1 MeV). Test parameters for both control and irradiated specimens included strain rates from 3 × 10−4 to 1 sec−1 and temperatures from 22 to 927°C. The results of these tests were correlated with a rate-temperature parameter ( T ln A ϵ ) to provide a concise description of material behavior over the range of deformation conditions of this study. The yield strength of the subject materials was significantly increased by decreasing temperature, increasing strain rate, and increasing fluence. Ductility was significantly reduced at any temperature or strain rate by increasing fluence. Cleavage fractures occurred in both unirradiated and irradiated specimens when the yield strength was elevated to the effective cleavage stress by temperature and/or strain rate. Neutron irradiation for the conditions of this study increased the ductile-to-brittle transition temperature of TZM by about 215°C and tungsten by 150°C.

Swelling in neutron irradiated tungsten and tungsten-25 percent rhenium

Swelling in neutron irradiated tungsten and tungsten-25 percent rhenium

Superlattice of voids in neutron-irradiated tungsten

Transmission electron microscopy of tungsten irradiated at about 550°C to a fast neutron fluence of approximately 1×1022 neutrons cm−2 reveal the formation of a superlattice of voids. The void lattice parameter is 195 Å and the lattice parameter to average void size is about six. Both the formation of the superlattice and the temperature dependence of the lattice parameter in tungsten, a material considered to have isotropic elastic properties, is in contradiction to present theories based on the elastic anisotropy of the host metal.

Recovery of Low-Temperature Fast Neutron Irradiated Tungsten

The recovery after fast neutron irradiation at about 10°K of W with various treatments is investigated by electrical resistivity measurements and the recovery structure is determined from 15°K to 420°K. The stage I below 90°K of W consists of at least five substages after fast neutron irradiation. Radiation doping enhances the amount of recovery in stages I and II. In W , the effect of doping does not vanish when doped is annealed at 400°C, where stage III is completed. This is contrast with Mo where the enhancement vanishes when doped specimen is annealed above stage III.

Vacancy damage in heavy ion and neutron-irradiated tungsten

Abstract Vacancy damage produced by heavy ion, and neutron irradiation, of tungsten, at ambient temperatures, has been examined by both field ion and electron microscopy. Field ion microscopy revealed defects in the form of single vacancies, irregularly shaped voids (compact clusters) containing up to 300 vacancies, ‘dispersed’ clusters, and both perfect and imperfect dislocation loops with Burgers vectors ½a〈111〉 and ⅓a〈112〉, respectively. At least 80% of the damage consisted of small clusters, each containing less than ten vacancies. The only defects that were readily resolved by electron microscopy were dislocation loops, with b = ½a〈112〉 or ½a 〈 111 〉. Less than 10% of the total concentration of vacancies produced by irradiation was in the form of dislocation loops. Similar types of damage were produced by both heavy ion and neutron irradiations. Equivalent densities of damage were produced by doses 1 × 1017 (150 kev) Hg+/m2 and 5.42 × 1023 nf/m2.

Comments on Dislocation Loops in Neutron Irradiated Tungsten by J. M. Galligan and T. Oku

Comments on Dislocation Loops in Neutron Irradiated Tungsten by J. M. Galligan and T. Oku

A study of radiation damage in tungsten—I

Some direct measurements of the kinetics of vacancy removal in neutron irradiated tungsten are presented. These measurements show that vacancies are mobile in stage IV recovery of this material. A comparison of this experiment with the self diffusion measurements of Andelin, Knight and Kahn and the quenching measurements of Schultz, shows that the activation energy for self diffusion is consistent with a vacancy mechanism.

Point defect recovery in neutron-irradiated tungsten at 0.31 Tm

The results of some resistivity studies of the recovery of defects in neutron-irradiated tungsten provide additional evidence for the recovery of monovacancies at 0.31 T m.