Nuclear Materials and Energy最新文献

Li₄SiO₄ and Li₂TiO₃ are granular materials in the pebble bed of fusion reactors, and their physical parameters and mechanical properties directly affect the working status and structural design of the pebble bed. The force chain can be employed to intuitively describe the mechanical properties of the pebble bed in the particle aggregate. Based on the discrete element method, we conducted a numerical study on the evolution characteristics of the force chain of Li₄SiO₄ and Li₂TiO₃ pebble beds under biaxial compression and explored the friction coefficient and pebble bed effect of the aspect ratio on the force chain distribution. The results revealed that as the particle friction coefficient increased, the force chain tended to be isotropically distributed, indicating that the friction mechanism was more conducive to a uniform distribution of the force chain. During compression, the average coordination number of the pebble bed increased with the friction coefficient. The Li₂TiO₃ particles had larger gravity than the Li₄SiO₄ pebble bed therefore, the Li₂TiO₃ pebble bed accounted for more force chains in the vertical direction. When the aspect ratio of the pebble bed was less than 0.5 or greater than 2.5, the distribution of force chains exhibited strong anisotropy. Conversely, when the pebble bed aspect ratio ranged between 0.5 and 2.5, the distribution of force chains tended towards an isotropic trend. Moreover, the number of force chains in each direction varied with changes in the aspect ratio of the pebble bed, characterized by a high concentration at the edges and a lower concentration in the middle. The results can provide an in-depth understanding of the force chain distribution and evolution characteristics in the pebble bed and provide a theoretical basis for designing and analyzing tritium breeding pebble beds.

A new Divertor Tokamak Test (DTT) facility is currently being built in Italy to investigate different divertor configurations under different plasma scenarios. The divertor and in particular its target, consisting of Plasma Facing Units (PFUs), is exposed to high heat loads due to plasma fluence. In this paper, the structural integrity of the PFU is evaluated for a reference Single Null (SN) divertor configuration under three different plasma scenarios. A comprehensive structural integrity analysis has been carried out in three stages. In the first stage, computational fluid dynamic (CFD) analyses of the divertor’s PFU with a cooling channel were performed at three different heat loads corresponding to three plasma scenarios. The temperature fields calculated by the CFD analyses were then used as input for the second stage, in which thermo-mechanical simulations were performed to predict the stresses and displacements in the PFU. Due to the high local heat loads, high stresses or even yielding are expected in the PFU’s structural components. Therefore, in the third stage, the structural integrity of critical cross-sections has been verified using the Structural Design Criteria for In-vessel Components (SDC-IC). It has been demonstrated that structural components of the PFU are able to withstand the expected loads, although some non-structural components experienced yielding while not exceeding the critical values.

The divertor must simultaneously withstand unprecedented high heat fluxes of up to 20 MW/m² and high-energy neutron irradiation of 14-MeV during fusion reactor operation. Accordingly, it is necessary to simultaneously meet the excellent heat dissipation capacity of the divertor and maintain good material performance in harsh neutron irradiation environments. The flat-type divertor demonstrates better heat transfer performance compared to monoblock divertor. Nevertheless, under the condition of a heat flux of 20 MW/m², the heat transfer and thermal fatigue performance of flat-type divertor using advanced materials are still unidentified. In this study, we conducted a thorough analysis of the heat transfer capabilities and thermal fatigue characteristics of potassium-doped tungsten (KW)/Cu/Oxide dispersion strengthened copper (ODS-Cu)/reduced activation ferritic/martensitic (RAFM) divertor mockup adopts hypervaportron (HV) structure by numerical simulations combined with experiments. The numerical results indicate that the flat-type KW/Cu/ODS-Cu/RAFM divertor mockup expresses excellent heat transfer capacity. The contact area between the edge of the fins and the bottom surface of the ODS-Cu heat sink induces a significant increase in water flow velocity to ∼15 m/s. The peak temperature of loaded KW surface is only ∼985 °C under the flow rate of 5 t/h and inlet temperature of 20 °C. During the high heat flux tests, the prepared flat-type divertor mockup successful endured 1000 cycles of 20 MW/m² with the peak temperature of 883 °C, and the surface temperature experienced a fluctuation of 2.4 % during the thermal fatigue tests. This study can provide a strong data reference and technical support for the development of fusion reactors, and is of great significance in advancing the commercialization of fusion energy.

The interaction between ions should be greatly modified under electronic excitation states, subsequently altering the interactions between materials. We perform a series of first-principles calculations to predict the solution and diffusion behaviors of interstitial hydrogen (H) in tungsten (W) under various electronic excitations. Qualitatively, the solution, diffusion, and trapping behaviors of H in W under various electronic excitation states are basically consistent with those in the ground state. However, it can be found that the solution energy and the migration energy barrier of H decreases as increasing the electronic temperature of system. The Pearson correlation coefficient study shows that there exists a perfect negative correlation between the lattice constant of W and H solution energy induced by lattice distortion. Besides, electronic excitations also make the binding energy of multiple H atoms decrease. That is, when the same number of H atoms are added to the vacancy, the binding energy decreases with increasing the electronic temperature of system. Based on these calculation results, we can infer that electronic excitations make dissolved H atoms more active in W system. This may, to some extent, allow dissolved H to migrate around and not aggregate so easily, thus reducing the production of H bubbles. Therefore, in quantitative terms, the electronic excited states have a certain effect on the H behavior in W.

The key system to know the neutron and gamma fields during the commissioning at IFMIF-DONES facility will be the STart-Up Monitoring Module (STUMM), which will include tens of very sophisticated and miniaturised detectors providing crucial information about the radiation fields. In particular, there will be MFCs (micro-fission chambers), ICs (Ionisation Chambers), SPNDs (Self-Powered Neutron Devices) and GTs (Gamma Thermometers).

The current state-of-the-art does not provide answers about the performance of the detectors in the very specific working conditions inside the IFMIF-DONES Test Cell, as a very high spatial density of detectors inside the STUMM vessel, an extremely harsh environment with neutron flux up to 5·10¹⁴n/cm²/s and the need of long mineral cables (>30 m) to drive the very weak current signals from the irradiation area to the closest electronics cubicles available in the facility.

Except for the extremely high neutron flux provided by IFMIF-DONES, all the other very critical conditions can be replicated by means of a 1:1 scaled prototype of the STUMM (STUMM-PROTO), which will be subjected to comprehensive testing and irradiation campaigns in order to increase the knowledge about the performance of the detectors above mentioned.

The present work shows the wide range of capabilities offered by STUMM-PROTO as well as the current status of its construction and some first notions about the experimental campaigns to be implemented.

Isolated tungsten nanocolumns (W-NCs) have been reported to exhibit a higher radiation resistance than coarse grained W samples, under radiation conditions similar to those that plasma facing materials would face in both magnetic and inertial confinement nuclear fusion approaches (MCF and ICF, respectively). This is so, because their ability to release He via the free surfaces and, the strong reduction of the sputtering yield under KeV Ar and D irradiation as well as, flattening of its angular dependence. The latter is very important for the divertor location in MCF.

In this work, we investigate the capabilities of sputtering (an easy control, environmentally friendly, versatile, scalable and low-cost technique) to fabricate isolated W nanocolumns. We study the influence of sputtering parameters (plasma power and deposition angle) on the morphology, density and microstructure of the deposited coatings. Results show that stable α-phase 3D isolated W-NCs are produced at low plasma power (<100 W) and high deposition angles (θ ≥ 75°).

Polycrystalline specimens of copper oxide superconductors YBa₂Cu₃O_7-δ(YBCO), Y_0.39Sm_0.30Eu_0.31Ba₂Cu₃O_7-δ(ME-REBCO), Y_0.18La_0.24Nd_0.14Sm_0.14Eu_0.15Gd_0.15Ba₂Cu₃O_7-δ (HE-REBCO) were irradiated with 1 MeV-He ions at room temperature. Magnetization measurement using superconducting quantum interference device (SQUID) and microstructure observation using transmission electron microscope (TEM) were conducted before and after irradiation. The improvement of properties in HE-REBCO and degradations in YBCO, ME-REBCO were suggested by SQUID measurement. HE-REBCO had smaller irradiation defects and volume fractions although disappearance of twin boundaries due to tetragonal transition occurred in peak-damage area as well as YBCO. These defects work as optimal flux pinning centers leading to improve properties in HEREBCO.

Coolant-facing alloys in a tokamak will face harsh environments due to the synergistic effect of flowing coolant, dynamic mechanical loading, ionising radiation, and tritium embrittlement. Reduced activation ferritic/martensitic (RAF/M) EUROFER-97 is the primary candidate as the structural alloy for the breeder blanket to be used in DEMO reactor. Under fusion-like conditions, coatings are required to improve both its corrosion properties and act as a tritium permeation barrier (TPB) layer. In this study, Al-base self-passivating coatings have been investigated as possible solution to provide protection against corrosion and act as TPB layer. The aim of this study was to assess the adhesion properties between different self-passivating Al-base coatings as electroplated and chemical vapour deposited (CVD) coatings on a RAF/M EUROFER-97 under strain. These studies showed the high ductility of pure electroplated Al-coatings, whereas both CVD coatings cracked under the applied strain.

In this work, Zr addition is proposed to refine the nanoparticle dispersion in an ODS RAF steel of composition Fe-14Cr-2W-0.3Zr-0.24Y (wt.%). Three batches of material are obtained using pre-alloyed atomized powder, where yttrium is directly introduced in the melt, and manufactured through three different processing routes. First route is based on the newly developed STARS route that aims to avoid subsequent mechanical alloying. The second route explores the impact of mechanical alloying in pre-oxidized powders. The third route uses mechanically alloyed powders without the pre-oxidation process. The ODS-powders were individually consolidated by hot isostatic pressing and later hot rolled. The obtained materials were characterized by small-angle neutron scattering (SANS) and X-ray absorption spectroscopy (XAS) techniques. SANS and XAS analysis point out the absence of oxide nanoparticles in the material based on the STAR route. SANS analysis confirms that the mechanically alloyed materials do exhibit the presence of nanoparticles. These are identified as Zr-O-rich nanoprecipitates by XAS and the calculated A-ratio by SANS is linked with the phase Y₂Zr₂O₇. Their radii are in the range of 3–3.6 nm. XAS results show that mechanical alloying minimizes the initial differences regarding the oxidation state between the ODS powders with and without pre-oxidation.

This research aims to investigate the application of the hot pressing (HP) sintering process in the fabrication of W-B-Fe-Cr-C advanced shielding materials and analyze the relationship between their microstructure and mechanical properties. The findings reveal that the HP sintering process can effectively produce reactive sintered boride (RSB) materials with high density and engineering application size. The results indicate that the residual stress in RSB-HP improves its bending strength to some extent. Residual stress and granular structure are both related to phase aggregation during sintering. Microstructural analysis has unveiled the granular organizational characteristics and the pathways of crack propagation within the RSB-HP. These discoveries hold significant importance for comprehending the potential of the HP sintering process in the context of nuclear fusion shielding material preparation and provide a scientific foundation for further optimizing the sintering process and performance of RSB materials.