Nuclear Materials and Energy最新文献

The arc operating voltage and the mass/charge composition of the arc plasma were measured for W-fuzz samples with different fuzz thicknesses at different numbers of arc pulses. It was found that the formation of a fuzz structure led to a significant decrease in the arc operating voltage within the first few arc pulses. With an increase in the number of pulses, the operating voltage and the mean charge of tungsten ions in the arc plasma increased to values characteristic of conventional tungsten cathodes. The obtained experimental data were analyzed theoretically using an empirical cohesive energy rule and a magnetohydrodynamic model of the ionization processes occurring in a cathode plasma jet.

In recent years, medium and/or high-entropy alloys (M/HEAs) have attracted attention due to their excellent mechanical properties, positioning them as potential candidates for structural materials in advanced nuclear reactors. This study investigated the effect of He injection amount on the microstructural evolution in Cr_0.8FeMnNi MEA and 316L stainless steel. Helium atoms were injected, followed by Fe²⁺ irradiation. Transmission electron microscopy was used to characterize the irradiation-induced defects. The primary irradiation-induced defects observed across all He injection amounts were black dots; no voids were seen under any irradiation conditions. In He-injected Cr_0.8FeMnNi, the number density of black dots was significantly lower than in non-He-injected Cr_0.8FeMnNi, while the average size of black dots increased with higher He injection rates. As the He injection amount increased, the binding energy of the self-interstitial atom (SIA) to the He-V complex decreased. This reduction facilitated the nucleation and growth of numerous SIA clusters within the matrix, consequently increasing the number density of black dots. The change in the number density of black dots in Cr_0.8FeMnNi was much smaller than that observed in 316L. This could be ascribed to the enhanced self-healing process caused by the diversity of point defects and higher lattice distortion in Cr_0.8FeMnNi.

X-Point Radiator (XPR) regimes have been obtained in WEST tokamak experiments with nitrogen seeding during the experimental campaigns of 2023 and 2024. These experiments showed the formation of a stable toroidal radiating ring near the X-point, similar to observations in other devices such as JET, ASDEX-Upgrade, TCV, and COMPASS. In WEST, the onset of this regime is associated with a sharp transition of the divertor plasma from hot to cold and dense conditions, with increased particle fluxes, indicating that the plasma is not detached. At the same time, core conditions are significantly improved. These scenarios were successfully controlled in WEST using an interferometry line-of-sight passing through the X-point. Interpretative modeling of these discharges with the SOLEDGE3X-EIRENE code reveals that a physics mechanism needed to stabilize WEST nitrogen XPRs is not present when driving the simulations at constant power. On the contrary, a stable XPR can be obtained by increasing the power injected at the time of the XPR onset to represent the reduction of W contamination, highlighting the need to describe the plasma dynamics and go toward integrated core–edge simulations.

Reactor relevant fusion devices will use tungsten (W) for their plasma facing components (PFCs) due to its thermomechanical properties and low tritium retention. However, W introduces high-Z impurities into the plasma, degrading its performance. Different wall conditioning methods have been developed to address this issue, including coating of W PFCs with layers of low-Z material. Wall conditioning by boron (B) powder injection using an impurity powder dropper (IPD) is being studied in WEST. Two series of experiments were conducted since the installation of the new ITER grade full W divertor. During the first series in 2023 ∼ 1 g of B powder was injected in total at a maximum rate of ∼ 58 mg/s, both of which are three times greater than respective values in the initial WEST powder injection experiments. The second series of experiments included injection of B and BN powders for comparison of their effects on plasma performance. The presence of an instantaneous conditioning effect is suggested by visible spectroscopy measurements of low-Z impurity lines and a rollover of total radiated power past an injection rate of ∼ 20 mg/s was observed. Presence of B coating layer formation is supported by the evolution of the average radiance of visible lines of B, W and oxygen (O). To understand B transport, an interpretative modeling workflow is employed, utilizing the SOLEDGE-EIRENE fluid boundary plasma code and the Dust Injection Simulator (DIS) code. Parameters like B perpendicular diffusivity and recycling coefficients are varied to match experimental results to see if the initial assumption of B sticking to the PFCs immediately after the contact with the wall is adequate for correctly modelling its distribution on the PFCs.

Arcing is a mechanism of dust production and erosion in fusion devices. The masses and velocities of tungsten (W) particles produced by arcing is of special interest given that W is used as wall material in current experimental devices and is planned for use in new machines such as ITER. This work presents a new technique for evaluation of the parameters of particles produced by vacuum arcs in laboratory conditions based on direct observation with a high-speed video camera. Tracking of particle trajectory provides a measurement of velocity and angle of emission. Additionally, the emitted thermal radiation of the particles is measured and its evolution over time is compared with a model of its cooling in order to obtain a measurement of size and temperature at the moment of emission for each individual particle. Surprisingly, temperature measurements reveal the production of initially solid particles. The newly established video technique allows automated measurement of a high number of particles in order to obtain distributions of the particle parameters.

Irradiation hardening is one of the service performance concerns for tungsten in fusion reactors. This work investigated the effect of deuterium (D) plasma exposure and helium (He) ions irradiation on the hardening behavior of the same tungsten sample using nanoindentation. The results demonstrate that the hardness of tungsten increases after D plasma exposure, He ions irradiation, and synergistic irradiation of He-ion and D-plasma. In general, the degree of hardening is He-ion + D-plasma irradiation, individual He-ion irradiation, and D plasma exposure in descending order. D plasma exposure results in an increase in hardness of more than 10 %, which is attributed to the pinning of dislocations by D plasma-induced defects and D-defect complexes (clusters). In the case of the He-ion irradiated tungsten, a large number of defects such as He nanobubbles induced by He-irradiation result in a 70 % increase in hardness. The superposition effect on the hardening of tungsten by D plasma exposure after He ions irradiation was observed, which essentially remains an increase in hardness due to D plasma exposure. This implies that the degradation of mechanical properties caused by D plasma exposure on tungsten will not be overlapped by other particle irradiation. Moreover, He bubbles in tungsten remain stable and grow slightly after annealing at 1173 K, resulting in a limited decrease in hardness.

The vacancy and helium effects on eight low-Σ symmetric tilt grain boundaries (STGBs) of Ni were investigated by first-principles calculations. The simulations demonstrate that vacancies in Ni matrix could diffuse easily to the grain boundary and helium atoms are quite stable at the grain boundary vacancy sites. Vacancy accumulation at high-energy STGBs could enhance the binding strength while He defects are generally detrimental to grain boundary binding and interstitial helium defects can initiate more severe grain boundary cracking comparing to vacancy-bind-helium. Further electron charge analysis suggested the influence of grain boundary binding greatly relies on the grain boundary atomic structure and the interaction between Ni-d and He-p states. The polarized charge transfer induced by He helium occupation was predicted to be detrimental to the binding strength, which explains the experimentally-observed helium bubbles and helium-induced GB embrittlement.

Compared to CuCrZr, oxide dispersion strengthened copper alloy (ODS-Cu) exhibits higher stability of properties under irradiation and exposure to elevated temperatures, demonstrating broad application prospects in divertor components. Oxygen-free high thermal conductivity copper (Cu-OFHC) has been frequently employed as an interlayer between W and Cu-based alloy in the fabrication of W/Cu divertor components. This study investigates the effect of joining temperature together with the addition of a Ni interlayer on the interface microstructure and mechanical properties of the Cu-OFHC/ODS-Cu joints. As the joining temperature increased from 680 ℃ to 900 ℃, the interface bonding ratio of the Cu-OFHC/ODS-Cu joints improved from 40.4% to 95.8%, and the impact toughness increased from 23.4 J/cm² to 133.1 J/cm². With the addition of a Ni interlayer, the interface bonding ratio increased from 40.4% to 90.3%, and the impact toughness improved from 23.4 J/cm² to 122.5 J/cm². Increasing the joining temperature or adding a Ni interlayer effectively reduced interfacial voids, enhanced the interface bonding ratio, and consequently improved the impact toughness of Cu-OFHC/ODS-Cu joints. Then, the W/Cu/ODS-Cu monoblock mock-ups with good interfacial bonding were successfully fabricated under two conditions: at a joining temperature of 900 ℃ without an interlayer and at 680 ℃ with a Ni interlayer. These results provide a fundamental understanding for achieving high-quality Cu-OFHC/ODS-Cu joints and offer technical support for the engineering preparation of W/Cu/ODS-Cu components in future fusion devices.

This paper constructs four alloy structural models of W₁₅Cu₁, W₁₄Cu₂, W₁₃Cu₃ and W₁₂Cu₄ as research objects. Based on phonon spectra analysis, the structure of W₁₂Cu₄ is unstable. Thus, the mechanical and thermodynamic properties for pure W, W₁₅Cu₁, W₁₄Cu₂ and W₁₃Cu₃ are studied using first-principles methods. The results show that the bulk modulus of W-Cu alloys exhibits the characteristics of high-temperature strength alloys. In particular, the bulk modulus of W₁₄Cu₂ and W₁₃Cu₃ increases rapidly with temperature increasing; Meanwhile, The W-Cu alloys at high temperatures exhibit low expansion behavior. The coefficient of thermal expansion of W₁₄Cu₂ at 1290 K starts to be lower than that of W, and decreases with temperature increasing. At 1370 K, the coefficient of thermal expansion of W₁₃Cu₃ begins to be lower than that of W₁₄Cu₂, which also decreases with temperature increasing. The W-Cu alloys at high temperatures exhibit excellent phonon thermal conductivity. The phonon thermal conductivity of W₁₄Cu₂ at 2230 K transcends pure W and then maintains a thermal conductivity of about 5.55Wm^-1K⁻¹. The phonon thermal conductivity of W₁₃Cu₃ at 2140 K begins to transcend W₁₄Cu₂, and also exceeds pure W at 2180 K, and increases rapidly with temperature increasing. At 0 K, the addition of Cu to W reduces the mechanical strength but increases the ductility.

Ensuring subcriticality over storage duration, even in accident condition, is one of the important requirements for dry cask designed to store spent fuel from nuclear reactor. To comply this requirement, neutron absorbers are incorporated in the spent fuel basket to decrease the neutron multiplication factor and reactivity. The objectives of this study are to prepare Cd/316L composite for neutron absorber material by powder metallurgy and determine the neutron shielding performance of the obtained composite. Powder mixtures of SS316L with various Cd contents were milled using high energy milling, followed by pressing and sintering at 450 °C and 700 °C. The analyses of the samples were carried out using XRD, SEM-EDS, XRF, and neutron transmittance test. Additionally, the neutron transmittance was also calculated using MCNP5. The results show that transformation of austenite to α-ferrite occurred during powder milling process which remained in the sample after sintering at 450 °C. Transformation of α-ferrite in the milled sample to austenite occurred after sintering at 700 °C. Cd in the samples were predominantly oxidized to form CdO phase during sintering process at both temperature conditions. The loss of Cd content in the Cd/SS316L samples after milling process was approximately 10 %. The Cd content further decreased by approximately 10 and 50 % after sintering at 450 and 700 °C, respectively. Neutron transmittance test shows that the transmittance value for thermal neutron decreased with the increased of Cd in the Cd/316 sample. A good agreement has been obtained between the results from experiment and MCNP calculation.