Neutron Irradiated Tungsten Research Articles

The effect of microstructure on recovery and recrystallization after annealing of two neutron irradiated ITER specification tungsten

The study of the recrystallization resistance of engineering grade tungsten (W) prior to and after neutron irradiation as well as subsequent thermal annealing provides valuable insight into neutron induced defect interactions and their kinetics as well as their correlation with recrystallization mechanisms. The outcome of such investigations would be of assistance for appropriate grade selection for a fusion reactor and the design of healing processes enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification W, forged bar and cold-rolled plate, were neutron irradiated to a dose of 0.18 dpa at 600 °C in the Material Test Reactor BR2, Mol, Belgium, and subsequently isochronally annealed from 700 to 1550 °C for 24 h with a 50 °C step with two non–irradiated counterparts. X-ray diffraction, optical microscopy and positron annihilation lifetime spectroscopy, combined with Vickers Hardness are used to correlate the recrystallization process and recovery mechanisms. It is found that the non-irradiated plate has a higher recrystallization resistance compared to that of the bar, with the former requiring 100 °C higher temperature for 24 h annealing time to recrystallize. Despite the irradiation induced dislocation loops and voids being of similar sizes and densities for both grades, the different initial microstructure affects the recrystallization process upon annealing, delaying it by 50 °C on the plate. Further in the plate a complete void dissolution is observed before recrystallization, which is not the case for the bar, for which the two effects occur at the same temperature.

Direct observation of voids decorated with transmuted rhenium atoms in neutron-irradiated tungsten by correlative use of TEM and APT

The microstructure of tungsten (W) irradiated by neutrons up to 1.54 displacements per atom at 750 ℃ in the fast-neutron experimental reactor Joyo, where the thermal neutron ratio is low, was characterized by observing same needle specimen by both transmission electron microscopy (TEM) and atom probe tomography (APT). TEM showed nanoscale voids with a high number density in the matrix, while APT showed nanoscale rhenium (Re)-enriched clusters with a similar number density in the matrix. By comparing the TEM image and APT map in the same area of the same needle specimen, a one-to-one correspondence was clearly obtained between the voids observed by TEM and the Re-enriched clusters observed by APT. Furthermore, the size distributions for the voids and Re-enriched clusters were almost the same. Thus, it was found that all voids were decorated with transmuted Re atoms in the (initially pure) W neutron-irradiated in Joyo where the thermal neutron ratio is low.

Atomically-informed dislocation dynamics model for prediction of void hardening in irradiated tungsten

Nanometric voids are one of the fundamental defects generated by neutron irradiation in fusion environment, which impede dislocation motion and cause hardening of plasma facing tungsten. Accurate prediction of voids hardening is crucial for safety evaluation of irradiated tungsten. In this paper, we develop a void-dislocation interaction model within the framework of dislocation dynamics (DD) and implement it in open-source DD software ParaDiS. We use this model to study the hardening effects of randomly-distributed voids with different sizes and number densities. The model for short-range interaction during dislocation-void contact, especially the dislocation movement on void surface, is informed and calibrated by atomistic simulations. Meanwhile, the long-range elastic interaction between dislocation and void is efficiently computed by the approach developed in our recent work (Wu et al., 2022) with Eshelby's equivalent inclusion method. Our DD simulations show that the average resistance by voids with various sizes is proportional to the 2/3 power of total number density. We also find that a modified Friedel-Kroupa-Hirsch (MFKH) model can quantify the hardening effects of nanometric voids, where the hardening coefficient depends on an effective average void diameter that accounts for the size distribution of voids. This MFKH model extends our previous work (Wu et al., 2022) that addresses the hardening of small vacancy clusters where dislocations simply cut through them, which is essentially an elastic perturbation to the dislocation dynamics. The proposed MFKH model can predict both small vacancy clusters and nanometric voids hardening effects with only difference in size dependent hardening coefficient. We further verify the MFKH model by comparing its quick predictions of the hardening effects by randomly-distributed nanometric voids with experiments on twelve neutron-irradiated tungsten samples.

Surface chemistry of neutron irradiated tungsten in a high-temperature multi-material environment☆

Surface chemistry of neutron irradiated tungsten in a high-temperature multi-material environment☆

Defect evolution of neutron irradiated ITER grade tungsten after annealing

The microstructural evolution of neutron irradiated tungsten (W) after annealing and its correlation with the corresponding mechanical properties provides valuable insight on the defect interactions and their annihilation processes. This would result in the identification of recovery mechanisms, leading to the design of healing processes and thus, enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification forged W bar were neutron irradiated to a dose of 0.2 displacements per atom (dpa) at 600 °C in the Belgian Material Test Reactor (BR2) and subsequently annealed at 800 and 1000 °C. The evolution of the irradiation induced defects after annealing has been assessed by transmission electron microscopy (TEM), positron annihilation lifetime spectroscopy and electrical resistivity and its effect on the mechanical properties are discussed in terms of Vickers hardness. Neutron irradiation results in the formation of dislocation loops and voids. TEM observations show an increase in the size of both defect types after post irradiation (PI) annealing at both temperatures, accompanied by an initial increase in their density after PI annealing at 800 °C, followed by a subsequent decrease after PI annealing at 1000 °C. The presence of TEM-irresolvable defects of both types is revealed in the as-irradiated state, which is evidenced by the evolution of Vickers hardness, resistivity and positron lifetimes and their correlation after the PI annealing. In the as-irradiated state, as well as after PI annealing at both temperatures, the hardening is dominated by voids.

T-REX: Numerical tool for tungsten damage assessment for DEMO

One of the R&D focus in the European fusion energy program is to establish a physical and technological basis for reliable power exhaust during entire operational situations of a DEMOstrational power plant. Following the design adopted for ITER, the baseline Plasma Facing Component for DEMO divertor is made of tungsten as armor material. The tungsten divertor target are the most thermally loaded in-vessel components in fusion reactor. The PFCs lifetime is affected by material degradations under the different loadings including High Heat Flux (HHF) and neutron irradiation. It has been reported a loss of mechanical properties due to softening (restoration/recrystallization). Latest finite element modeling developments allow to estimate component lifetime under HHF loading taking into account the progressive mechanical properties change due to softening. We propose here to take into account the influence of thermal and neutron loadings on the tungsten mechanical properties change for PFC lifetime assessment. For this, up to date mechanical properties of tungsten, softened tungsten, and neutron-irradiated tungsten are used as input data in the tool named T-REX (for Tungsten-Response based on EXperimental data). This tool is implemented as a user subroutine in a finite elements code. The influence of the neutron irradiation is studied as function of the dpa dose, the incident heat flux intensity and the current defined divertor component geometries (ITER and DEMO). With the study performed, it is shown that the geometry may be optimized to reduce the damage induced by plasticity. The numerical results also highlights that the plastic strain increment obtained after each thermal cycle reduces with the dpa dose.

Effect of rhenium addition on deuterium retention in neutron-irradiated tungsten

The effects of rhenium (Re) addition on deuterium (D) retention in neutron-irradiated tungsten (W) were investigated. Pure W and W-5Re (5 at.%) alloy samples were irradiated with neutrons at High Flux Isotope Reactor using MFE-RB-19 J capsule. The sample temperature and the damage level were 864 K and 0.35 dpa for pure W and 792 K and 0.26 dpa for W-5Re alloy. A portion of the samples was exposed to D plasma at Tritium Plasma Experiment at Idaho National Laboratory at 823 K to a fluence of 5 × 1025m−2. Vacancy-type defects in neutron-irradiated samples were examined using positron annihilation spectroscopy (PAS); D retention after plasma exposure was evaluated by thermal desorption spectrometry (TDS).TDS measurements revealed that D retention in the neutron-irradiated W-5Re alloy was similar to that in the unirradiated W sample, whereas a significant increase in D retention was observed in neutron-irradiated W. Thus, Re addition significantly suppressed the increase in D retention after neutron irradiation. This effect was attributed to the suppression of vacancy-type defect formation, as confirmed by PAS.

Positron annihilation spectroscopy investigation of defects in neutron irradiated tungsten materials

The identification of defects in neutron damaged materials is essential for elucidating the correlation between the microstructure and the properties of a material. Positron annihilation spectroscopy (PAS) is very sensitive in open volume defects and thus a useful tool for the investigation of the radiation damage in matter. Tungsten is a critical material for the first wall and divertor of fusion reactors. In the current work, the evolution of the open volume defects in tungsten (W) materials neutron irradiated to 0.12 displacements per atom (dpa) and in the temperature range from 600 to 1200 °C in the Belgian Material Test Reactor BR2 is investigated by positron annihilation lifetime and coincidence Doppler broadening spectroscopy. Three tungsten grades were studied: W(100) single crystal, ITER grade forger bar and heavily deformed “cold”-rolled sheet. PAS results show that the neutron irradiation results in the formation of dislocations and voids of size larger 1 nm at all irradiation temperatures and in all W grades. The dislocation and void density decreases with increasing irradiation temperature. Moreover, the void size increases with the increase of the irradiation temperature.

Penetration of deuterium into neutron-irradiated tungsten under plasma exposure

Hydrogen isotope trapping at lattice defects in neutron-irradiated tungsten (W), a leading candidate as plasma facing material, is an important problem determining tritium (T) inventory in a vacuum vessel of a future fusion reactor. In this study, W samples were irradiated with neutrons in the Belgium Reactor 2 at 563 K to 0.06 or 0.016 displacement per atom (dpa). After characterizing defects by positron lifetime measurements, deuterium (D) penetration under exposure to D plasma was examined at 563–773 K. Positron lifetime showed the presence of dislocations, monovacancies and relatively large vacancy clusters. These defects trapped D atoms with different values of binding energy. Dependence of D retention on plasma exposure temperature and damage level indicated that the concentrations of weak traps with smaller binding energy increased more significantly with damage level than those of strong traps.

Application of sub-miniaturized bending tests to extract tensile properties from neutron-irradiated metallic alloys

In this work, we demonstrate the applicability of sub-miniaturized three-point bending (3PB) testing for the extraction of tensile plastic properties of metallic materials, such as yield stress and work hardening rate. The approach is developed and validated on the example of tungsten. For this, dedicated finite element method (FEM) simulations are performed to reveal the correlation between the flexural stress-strain response in 3PB tests and yield stress in tensile tests. Furthermore, a dedicated inverse FEM procedure utilizing rate-sensitive isotropic physical model is developed to extract both yield stress and work hardening rate from the available set of bending tests. Both FEM-based approaches are first benchmarked on unirradiated materials and then validated using neutron-irradiated tungsten grades. The accuracy of the tensile properties extracted using both methods is discussed along with the estimation of a number of 3PB tests required to reach a target precision. The applicability of 3PB geometry for the extraction of tensile properties from unirradiated bending tests below ductile-to-brittle transition temperature (DBTT) is also addressed.

Numerical analysis of deuterium migration behaviors in tungsten damaged by fast neutron by means of gas absorption method

Deuterium retention behavior in tungsten damaged by fast neutrons at high temperatures (0.43 dpa at 918 K and 0.74 dpa at 1079 K) and 6.4 MeV Fe2+ (0.3 dpa at R.T.) were investigated to evaluate the tritium retention property of fusion reactor divertors. A deuterium gas absorption method was carried out to avoid additional damage that may be induced by plasma exposure, then, deuterium retention and desorption behaviors were investigated quantitatively by means of thermal desorption spectroscopy and the following simulation code. The deuterium desorption spectra for tungsten samples were analyzed by the numerical code which includes the elementary steps of hydrogen isotope migration processes including diffusion, trapping, detrapping, and surface recombination. The evaluated deuterium detrapping energy from the irradiation defects in neutron irradiated tungsten sample was larger than that in 6.4 MeV Fe2+ irradiated tungsten. It was suggested that the dominant deuterium trapping site in the neutron irradiated tungsten would be voids which was formed by the accumulation of vacancies during neutron irradiation under high temperature and long duration.

New insights into microstructure of neutron-irradiated tungsten

The development of appropriate materials for fusion reactors that can sustain high neutron fluence at elevated temperatures remains a great challenge. Tungsten is one of the promising candidate materials for plasma-facing components of future fusion reactors, due to several favorable properties as for example a high melting point, a high sputtering resistivity, and a low coefficient of thermal expansion. The microstructural details of a tungsten sample with a 1.25 dpa (displacements per atom) damage dose after neutron irradiation at 800 °C were examined by transmission electron microscopy. Three types of radiation-induced defects were observed, analyzed and characterized: (1) voids with sizes ranging from 10 to 65 nm, (2) dislocation loops with a size of up to 10 nm and (3) W–Re–Os containing σ- and χ-type precipitates. The distribution of voids as well as the nature of the occurring dislocation loops were studied in detail. In addition, nano-chemical analyses revealed that the σ- and χ-type precipitates, which are sometimes attached to voids, are surrounded by a solid solution cloud enriched with Re. For the first time the crystallographic orientation relationship of the σ- and χ-phases to the W-matrix was specified. Furthermore, electron energy-loss spectroscopy could not unambiguously verify the presence of He within individual voids.

Mechanical properties of neutron-irradiated single crystal tungsten W(100) studied by indentation and FEM modelling

The aim of this work is to investigate the effect of high-temperature neutron irradiation on the mechanical properties of single-crystal tungsten W(100) studied by mechanical testing and modelling. Neutron irradiation up to 0.12 displacements per atom was performed at 600, 800, 900 and 1200 °C in the Belgium material test reactor BR2 at SCK•CEN in Mol. The mechanical properties of the irradiated tungsten were assessed by impulse excitation and depth-sensing indentation measurements. The values of Young's modulus, shear modulus and Poisson ratio are not affected by the irradiation. The mean hardness of the irradiated tungsten increases by 48, 45, 46, and 34% after irradiation at 600, 800, 900 and 1200 °C, respectively. The results obtained by the instrumented hardness tests were used as input for the finite element method model, applied to deduce the contributions coming from the neutron irradiation defects to the plastic deformation upon the indentation process, which include the impediment of gliding dislocations by sessile defects and modification of surface morphology due to the inhibited strain hardening ability through the dislocation-defect interaction. The validation of the fitted constitutive laws was realized by the crystal plasticity finite element method (CP-FEM) model which was applied to simulate the indentation load-depth curves, with certain assumptions on the irradiation-induced microstructure, and to calculate the distribution of stress under the indenter to investigate the extension of the plastic zone in the process of the indentation.

Neutron irradiated tungsten bulk defect characterization by positron annihilation spectroscopy

Abstract Positron annihilation spectroscopy was used to evaluate the defects in neutron irradiated tungsten exposed at five different irradiation conditions. The variables in neutron irradiation included temperature, displacements per atom (dpa), and neutron spectrum. A set of W, Re, WRe, and WReOs control samples were used in assessing the data. Positron annihilation lifetime spectroscopy and coincidence Doppler broadening measurements revealed that samples irradiated at 500 °C had more vacancy clusters than samples irradiated at higher temperatures. This trend was observed despite some higher temperature samples having a significantly higher dpa. Positron lifetimes indicate these are divided into large (>40) and small (

Effects of Helium Seeding on Deuterium Retention in Neutron-Irradiated Tungsten

Neutron-irradiated tungsten (W) samples were exposed to helium (He)–seeded deuterium (D) plasmas using a linear plasma device called Tritium Plasma Experiment in order to investigate the synergetic effects of neutron and He irradiations on D retention in W. Exposure to nonseeded D plasma was also performed for neutron-irradiated and nonirradiated W samples for comparison. Deuterium retention in neutron-irradiated W after D plasma exposure was two to three times larger than that in W without neutron irradiation. Nevertheless, He seeding in D plasma resulted in a drastic reduction in D retention. The cross-sectional observation by transmission microscopy showed formation of He bubble layers with a thickness of 10 to 20 nm. There is a possibility that alpha particles in fusion plasma reduce tritium retention in neutron-irradiated plasma-facing components with W layers.

Re-evaluating ion and neutron irradiation data with calculated applicable arc-dpa

ABSTRACT To develop advanced nuclear materials, lots of irradiation experiments have been done to various materials in different conditions. Accurate quantification of the defects induced by irradiation is important for researches in nuclear materials, especially when evaluating the damage caused by different particles. An advanced model – athermal recombination corrected dpa (arc-dpa) is recently introduced. In order to apply arc-dpa on experimental data, overall arc-dpa in ion and neutron irradiated iron, copper and tungsten are calculated based on the arc-dpa function for single cascade and the calculated PKA energy spectra. Data of irradiation experiments are re-evaluated with our applicable arc-dpa. The use of arc-dpa on experimental data showed good fit, giving new insights into the evolution of irradiation damage done by different particles.

Dynamics of Hydrogen Isotope Absorption and Emission of Neutron-Irradiated Tungsten

Dynamics of Hydrogen Isotope Absorption and Emission of Neutron-Irradiated Tungsten

Advanced synchrotron characterization techniques for fusion materials science

Characterization methods capable of providing critical information across multiple structural length scales are essential in materials exposed to the extreme environments such as anticipated fusion power systems. Complementary techniques capable of uncovering the complicated microstructural irradiation-induced evolution are also important to verify and validate advanced computational models. To date, the primary microstructural tools informing such lower-length scale models have included analytical electron microscopy, positron annihilation spectroscopy, atom probe tomography, and small-angle neutron scattering. In this paper, we discuss the application of state-of-the-art synchrotron-based x-ray characterization methods in fusion material research. Specifically highlighted are opportunities in leveraging synchrotron-based techniques to address fundamental and applied materials science challenges at various length scales and in support of modeling efforts. Examples presented in this article include: a combined small angle x-ray scattering and x-ray diffraction study of transmutation-induced precipitation in neutron irradiated tungsten, and the identification of size and structure of nm-scale transmutation precipitates and voids; quantitative characterization of thermodynamically predicted minor precipitate populations in advanced reduced activation ferritic-martensitic steels through high energy x-ray diffraction; and a review of recent synchrotron-based studies dedicated to quantifying the radiation response of fusion relevant materials. The latter includes a pair distribution function analysis investigation of neutron irradiated SiC with insights into the different radiation response of the silicon and carbon sublattices, and a dose dependent decrease in the size of defect free material.

An atom probe tomography and inventory calculation examination of second phase precipitates in neutron irradiated single crystal tungsten

Tungsten is the prime candidate material for the plasma facing divertor system and first wall armor of future nuclear fusion reactors. However, the understanding of the microstructural and chemical evolution of pure tungsten under neutron irradiation is relatively unknown, in part due to a lack of experimental data on this topic. Here, single crystal tungsten has been irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory at a nominal temperature of 700–800 °C to damage levels of 0.1 and 1.8 displacements per atom (dpa). Inventory calculations of the neutron irradiation experiments have been used to track isotope generation and decay to inform atom probe tomography (APT) results in the determination of the transmutation-induced precipitate compositions. Furthermore, APT crystallography has been used to show the relationship between precipitates and matrix. The composition of the precipitates is shown to progress towards that of the σ-phase at the highest dose studied, with those precipitates lying along crystallographic planes similar to those of displacement-induced dislocations. This work also sets the framework for APT studies of materials that contain isotopic ratios far from those observed in the natural state.

Influence of radiation defects on deuterium permeation behavior in tungsten

Understanding the hydrogen isotope (H) permeation behavior in neutron irradiated tungsten (W) is highly desired for the accurate prediction of H behavior in W first walls in a fusion DEMO reactor. Contrary studies have been published where some indicated a relatively higher H permeability in the irradiated W whereas others showed a lower H permeability. In these studies the radiation defects were produced with ions (as surrogates for neutrons) and were concentrated in a shallow layer (<200 nm), their effects could be mixed up with surface effects during H permeation. Herein W specimens damaged to 800 nm by Au-ion implantation are adopted to investigate the effects of radiation defects on deuterium (D) permeation. Microstructural analyses with slow positron annihilation technique and transmission electron microscopy revealed that vacancy-type defects and dislocation loops were produced in the implanted W. Results from D2 gas-driven permeation measurements at 700–1000 K showed, surprisingly, radiation defects did not influence the D permeability. The phenomenon was illustrated from a thermodynamic perspective. Radiation-induced dislocations and/or monovacancies, possessing low H coverage fractions, do not affect permeability because their effects on effective diffusivity and solubility can fully cancel each other. Vacancy clusters, having a high fractional H coverage and owing to their strong trapping nature, will not come to effect during the steady-state permeation portion, thus do not contribute to permeability.