The development of materials for breeder blankets of fusion reactors requires a comprehensive understanding of their behavior under extreme operating conditions, including exposure to high temperatures, intense neutron irradiation, and contact with chemically active gaseous environments. In this context, metal beryllides, particularly Be12Ti and Be12Cr, are of significant interest and are considered as promising functional materials for breeder blanket applications.In this work, a phenomenological model of beryllide corrosion in water vapor is proposed based on the analysis of experimental data on the high-temperature corrosion of chromium beryllide (water vapor pressure ranging from 10 to 10,000 Pa and temperatures between 373 and 1100 К), as well as relevant literature data. The model assumes diffusion-controlled growth of a BeO oxide layer and enables the correlation of the experimentally observed kinetics of mass change and gas evolution with effective parameters describing beryllium diffusion through the oxide. It was found that a single set of model parameters provides an adequate description of all the experiments performed. The parameters, including the pre-exponential factor D0 (8·10-8 m2 s−1) and activation energy Ed (115 kJ·mol−1) in the Arrhenius expression for the diffusion coefficient, as well as C0 (1.75 mol·m−3), corresponding to the equilibrium concentration of mobile beryllium carriers in BeO at the metal-oxide interface, were reliably determined through fitting to the experimental data.
The increase of boron content leads to decreased toughness of high-boron stainless steel, which limits its application in the nuclear field. Hot isostatic pressing (HIP) sintering can effectively control the size and distribution of borides, improving plasticity and toughness. However, boron content significantly impacts microstructure and mechanical properties, particularly toughness. This study prepared high-boron steel samples with 2.06 wt% and 3.32 wt% boron using HIP, and systematically investigated the specific mechanism by which boron content affects toughness through microstructural observation, phase analysis, and comparative analysis of microstructure and texture before and after tensile deformation. Results show that when the boron content exceeds a critical proportion, the connection between boride grains causes an increase in boride grain size and phase fraction from 23.54% to 50.12%. The tensile strength remains comparable at 819 ± 1 MPa and 839 ± 23 MPa. However, the elongation decreases dramatically from 22.1 ± 2.6% to 1.2 ± 0.2%. In 2 wt sample, the interconnected austenite matrix accommodates coordinated deformation through dislocation cross-grain slip and grain rotation, resulting in ductile dimple fracture. In the 3.3 wt sample, the interconnected boride network restricts cross-grain dislocation motion and grain rotation, leading to brittle intergranular fracture with minimal plastic deformation. This study provides theoretical support for further optimizing the microstructure and mechanical properties of high-boron stainless steel materials.
In a fusion environment, tungsten (W) self-interstitial atoms (SIAs) are critical point defect in plasma-facing W materials, and their production and migration control the yield strength, ductility, radiation-induced growth of defects and nanobubbles, and structural stability. Although extensive theoretical simulations have predicted the generation and migration of W SIAs under hydrogen/helium plasma irradiation, direct experimental evidence confirming the existence of irradiation-induced W SIAs is still lacking. In this study, polycrystalline W samples are irradiated by a high-density hydrogen plasma under fusion-relevant conditions, followed by thermal annealing under oxygen atmosphere. Near-surface W atoms ejected by migrating W SIAs form surface W adatoms; these mobile adatoms diffuse across the W(110) surface and react with adsorbed oxygen species (O*) to form crystalline W18O49 nanowires, as verified by scanning electron microscopy (SEM), high-resolution transmission electron microscopy (HRTEM) and X-ray diffraction (XRD). Surface reaction rate model predicts that the density of W adatoms is strongly dependent on W temperature, and the surface diffusion of W adatoms contribute to the W growth, which affects the surface microstructure of hydrogen-irradiated W. This work explores a potential mechanism linking the formation of tungsten adatoms and W18O49 nanowires to radiation-induced W SIAs.
The corrosion behavior of construction with a lithium–lead eutectic (PbLi) is among the key issues for the application of a liquid PbLi blanket in a fusion device. In this study, the corrosion behavior of 310S stainless steel (SS) and Incoloy 800 alloy in liquid PbLi at 700 ℃ was investigated using a static immersion method. The weight loss, surface microstructure, and corrosion attack depth of the 310S SS and Incoloy 800 alloys were correlated with the elemental distribution. 310S SS and Incoloy 800 alloy exhibited non-uniform corrosion behavior, characterized by localized grain-boundary attack and porous corrosion structures, which may be associated with the selective dissolution of chromium (Cr) and nickel (Ni). Upon completion of the test, the oxide layer on the surface of the heating rod was disrupted, though some residue remained, which partially mitigated PbLi corrosion. The corresponding weight loss rates were 0.073 ± 0.005 g/m2/h for 310S and 0.118 ± 0.02 g/m2/h for Incoloy 800. Both materials appeared to suffer severe degradation after only 1000 h of exposure, and the corrosion penetration depth in Incoloy 800 ranged from 50 to 500 μm, which was notably greater than that observed for 310S stainless steel (about 45 μm). After exposure to liquid PbLi, a reduction in surface hardness was observed for both materials, and this decrease was non-uniform over the surface. The tensile properties of Incoloy 800 slightly decreased following PbLi exposure, whereas no significant variation was observed for 310S.
Carbon impurity sputtering and transport in the HL-3 tokamak are investigated using the time-dependent JOREK kinetic-fluid coupled framework with an updated sputtering model. Physical sputtering, chemical sputtering, and self-sputtering are treated simultaneously, with their yields evaluated dynamically from the evolving local plasma conditions. In the axisymmetric simulations, physical sputtering is localized near the strike point, whereas chemical sputtering extends over a broader target region. The sputtered carbon is subsequently transported along the divertor and scrape-off-layer flux tubes and exhibits a pronounced high-field-side/low-field-side asymmetry. During the simulated edge-localized-mode burst, the rapid increases in target temperature and deuterium ion flux strongly enhance and broaden physical sputtering, while chemical sputtering is suppressed near the strike point. Self-sputtering remains a secondary contribution under the conditions considered. These results demonstrate that the different sputtering channels respond differently to transient divertor conditions and should be treated self-consistently when modeling carbon source formation and transport in HL-3.
Many startups have been established in the world and there are many proposals on fusion reactors. Startups attract investment for the design and construction of fusion reactors. The development of high-temperature superconducting (HTS) tapes allows magnets to be demountable, making the vacuum vessel (VV) to be constructed independently of the magnet manufacturing process. The size of the VV is limited by the plasma facing material (PFM). High magnetic field reactors can generate more energy, but there is a limit to the heat load on the PFM. Since the VV is a nuclear boundary, welding must be flawless. To create a more effective design and clearly identify key problems, a targeted reactor design is necessary. A 300 MWe fusion reactor design was attempted and its size and structural components were visualized. HTS demountable TF magnets are expected, for it makes it possible to carry out VV sector-to-sector welding from both the inside and outside and it would reduce the welding flaws. Also, welding and non-destructive inspection of VV sector-to-sector were considered. This study describes some practical key issues on the design of a fusion reactor to be solved in near future, assuming 300 MWe fusion reactor.
Liquid lithium breeding blankets are a critical enabling technology for the operation of deuterium–tritium nuclear fusion power plants, providing with the necessary fuel for the reaction to take place. However, one can envision the liquid metal as providing with other critical functions such as neutron shielding and energy extraction. In this work we outline the path from conceptual design of a liquid metal wall subsystem intended to fulfill these various functions, to practical implementation. This implies an increase in Technology Readiness Level (TRL), a qualitative scale of the level of maturity of an industrial product. We define this metric as applied to Liquid Metal Plasma Facing Components and define a roadmap of successive experiments that will allow to progress several levels in this scale.Identified challenges include structural corrosion at fusion relevant temperatures due to the liquid metal flow, the use of active control to achieve an objective flow configuration using the electromagnetic forces resulting from injected electric current in the presence of a magnetic field, and proper distribution of lead pebbles intended to multiply neutron production. We describe recent numerical and experimental efforts that address these issues, and describe the roadmap of technology demonstrators necessary to build on the path towards a fully realized power plant.In this work we give an overview of several work streams that illustrate the multidisciplinary nature of the design and operation of these experiments. We first show results on high temperature static corrosion studies. We then propose a numerical methodology that allows to simulate injected current in a liquid metal flow and use it to design a magnetic levitation experiment. We describe a methodology that predicts the trajectories of solid pebbles, and find that design and control is possible via acting on the electromagnetophoresis force. Finally, we introduce a methodology for assessing the risk of free surface instabilities based on Linear Instability Analysis which accounts for system size.
The irradiation-induced swelling behaviour of oxide dispersion strengthened copper alloys was comparatively investigated by in-situ high-voltage electron microscopy under 1250 keV electron irradiation at 573 K. ODS-Cu, Zr-containing ODS-Cu (ODS-CuZr), annealed high-purity Cu, and mechanically alloyed Cu (MA-Cu) were examined to elucidate the role of microstructural defect sinks on void evolution. ODS-CuZr exhibited the highest swelling resistance, characterized by an extended incubation period, significantly reduced void size and number density, and an extremely low swelling rate. This enhanced performance is attributed to the high density of ultrafine Zr-containing oxide dispersoids, which act as sinks for irradiation-induced point defects and effectively increase the recombination and annihilation of interstitials and vacancies. In contrast, ODS-Cu showed earlier void nucleation and higher swelling than ODS-CuZr, while annealed Cu exhibited immediate void nucleation followed by void growth and coarsening due to its low sink quantity. MA-Cu displayed the most severe swelling behaviour, characterized by a high swelling rate and the absence of an incubation period due to the dense pre-existing nanocavities introduced during mechanical alloying. Combined with previous studies, a hierarchy of swelling resistance, ODS-CuZr > ODS-Cu > annealed Cu ≫ MA-Cu, can be summarised in the presented study. The findings highlight the effectiveness of Zr-containing dispersoids in enhancing the irradiation tolerance of Cu alloys.
Predicting alpha particle heat loads on first-wall components is essential for stellarator reactor design. We present a symplectic guiding-center integration method that traces fusion-born alpha particles from the plasma core to physical wall surfaces defined by CAD geometry. The approach uses Meiss–Hazeltine coordinates, which extend beyond the last closed flux surface, combined with segment–triangle intersection using the CGAL library. We compare the SIMPLE code with the LSODE-based BEAMS3D tracer for W7-X reactor conditions. Applying the method to the Proxima 5C quasi-isodynamic stellarator, we compare four wall representations at 250 mm offset: VMEC-extended coordinates, VMEC-to-wall interpolation, harmonic (map2disc) coordinates, and explicit STL mesh intersection. While total prompt-loss fractions are similar across methods (approximately 5% within 500 ms), the first-hit distributions depend on wall representation and mesh resolution. This sensitivity arises from the near-tangent incidence of prompt-loss particles, where small differences in effective wall orientation produce large changes in intersection location. Accurate prediction of divertor heat loads therefore requires explicit treatment of wall geometry rather than parameterized coordinate approximations. The collisionless guiding-center model does not resolve full-orbit, sheath, or material-response effects at impact.
W coatings have applications as a protective layer over the structural steel of the International Thermonuclear Experimental Reactor (ITER) temporary First Wall. In this work, a finite element simulation study of W/SS 316L functionally graded material (FGM) coatings was performed to determine the coating’s design parameters that reduce the thermal stress-discontinuity at the interface. Simulations considering only W coatings were performed initially to determine the ITER temporary First Wall equivalent thermal load conditions. Afterwards, the simulations of W/SS 316L FGM coatings were performed for varying the number of gradient interlayers while keeping the FGM thickness constant. Then, the FGM thickness was varied while keeping the number of gradient interlayers constant. In the simulations, the thermal load during the deposition and under the heat flux cycle of the ITER temporary First Wall were considered. Based on the simulation results, a 4-interlayer coating system with an FGM thickness of 0.8 mm between the top W coating and the steel substrate was recommended for the application.
GRCop-42 is a recently developed alloy with significant potential for use in nuclear thermal propulsion (NTP) and other extreme environment applications. The thermal transfer capabilities and mechanical strength at elevated temperatures make it highly desirable as a potential nozzle material. However, little is known about the radiation stability of this alloy, which is essential for any nuclear application. To rapidly screen the impact of displacement damage on microstructural and mechanical stability, this study uses self-ion irradiation (6 MeV Cu2+) coupled with Scanning Electron Microscopy (SEM), Energy-Dispersive X-ray Spectrometry (EDS), Electron Backscatter Diffraction (EBSD), Scanning Transmission Electron Microscopy (STEM), and Continuous Stiffness Measurement (CSM) nanoindentation. Irradiation damage was tailored to range from 0.17 dpa, the estimated damage during potential NTP operation to two orders of magnitude higher (17 dpa). EBSD observations showed a minimal average grain growth from 0.5 µm in the non-irradiated case to 0.9 µm after 17 dpa. Post irradiation STEM revealed that the precipitate dominated microstructure was heavily altered resulting in counteracting hardening and softening mechanisms and thus minimal change in measured hardness (240 MPa). Notably, the combined effects of grain coarsening and precipitate dissolution appear to offset radiation defect hardening, producing a dynamic microstructural response. These results are consistent with a strong role of competing hardening mechanisms in modulating the irradiation behavior of GRCop-42 and support its candidacy as a nozzle material for NTP related applications.
We report first-order reversal curve (FORC) measurements of neutron-irradiated Fe-1 wt% Cu alloy to elucidate the effects of irradiation-induced Cu precipitation on the FORC diagrams. The FORC distributions, calculated from the mixed second derivative of the FORCs with respect to magnetic field, exhibited a single peak along the coercivity and interaction field axes. With increasing neutron fluence, the peak initially shifted toward higher coercivity and interaction field, accompanied by peak broadening, and subsequently shifted toward lower coercivity at higher fluence. These results suggest that magnetic hardening and increased magnetic inhomogeneity due to nanoscale defect formation dominate in the early stage of irradiation, whereas recovery-related magnetic changes become dominant at higher fluence. Principal component analysis of the FORC diagrams showed that approximately 93% of the total data variance was captured by the first three principal components and the progression of magnetic-property changes during neutron irradiation could be effectively tracked using the principal component score plot. These findings indicate that FORC analysis provides a useful nondestructive approach for evaluating irradiation hardening in reactor pressure vessel steels associated with nanoscale defect formation.
In this study, Molecular Dynamics (MD) and Object Kinetic Monte Carlo (OKMC) simulations are adopted to explore the long-term defect evolution behavior of Fe-9Cr alloys. The effects of injected Fe ion concentrations (0–1000 appm/dpa) on irradiated microstructures are systematically evaluated at two typical irradiation temperatures of 573 K and 673 K. The results reveal that the modulation effects of injected self-ions are more prominent under high-temperature and high-dose irradiation conditions. At irradiation doses below 1 dpa, injected ions exert a minor influence on matrix defects. As the irradiation dose increases, such an effect becomes considerable: at 10 dpa, the density of large defect clusters varies by more than one order of magnitude under different injected ion concentrations. High-concentration injected ions substantially promote the nucleation and growth of self-interstitial atom (SIA) clusters while inhibiting the formation and evolution of voids. At high irradiation doses, high ion injection concentrations significantly raise the production ratio of <100>-type SIA loops. Furthermore, this work also elucidates the fundamental mechanisms by which injected ions affect the diffusion and segregation of solute atoms.
Austenitic stainless steels are widely used as structural materials in advanced nuclear systems but suffer from irradiation-induced swelling and mechanical degradation at high doses. In this study, the irradiation response of an austenitic stainless steel strengthened by uniformly distributed nanoscale carbide precipitates (ARES-6P) is investigated through combined experiments and simulations. Ion irradiation up to 251 displacements per atom (dpa) at 500 °C is employed to evaluate microstructural evolution and mechanical behavior. The reported cross-sectional transmission electron microscopy void data are used to quantify irradiation-induced void characteristics, while nanoindentation assesses mechanical properties. ARES-6P exhibits significantly reduced void swelling (∼3.1 % at 200 dpa), corresponding to approximately 13.3 % of that reported for conventional 316 stainless steel (∼23.3 %) under comparable conditions. Strain-gradient crystal plasticity models are developed for both unirradiated and irradiated states (151–251 dpa) and are calibrated and verified against nanoindentation and tensile results. The simulations indicate that irradiation leads to moderate hardening, increased yield strength, and reduced strain-hardening capacity. In addition, defect annihilation beneath the indenter is found to correlate strongly with the local von Mises stress field. These results demonstrate that nanoscale carbide precipitates effectively enhance irradiation resistance by suppressing defect clustering and void growth. This study provides insight into the design of radiation-tolerant austenitic steels for advanced nuclear material applications.
Titanium beryllide (TiBe12) is a candidate neutron multiplier material for breeding blanket concepts in fusion reactors. In this study, thermal cycling experiments were performed on TiBe12 blocks using induction heating and water cooling to investigate the response to transient thermal loading. Thermal cycling of a one-piece hexagonal TiBe12 block resulted in fracture during the first heating cycle, attributed to non-uniform induction heating associated with the closed-loop block geometry. In contrast, a block subdivided into 12 TiBe12 segments withstood 205 heating and cooling cycles between approximately 360 and 920 degrees C without macroscopic cracking or fracture. Additional experiments showed that TiBe12 segments tolerate heating rates up to 1.1-1.3 K/s and cooling rates of about 1.2 K/s without failure. The results demonstrate that block segmentation significantly improves thermal cycling stability and provide experimental data for validation of thermo-mechanical models of beryllide blanket components under reactor-relevant transient conditions.
To investigate the irradiation hardening mechanism involving dislocations and cavities interactions in tungsten, this study carried out the in-situ transmission electron microscopy observation under tensile test at 973 K. The obstacle barrier strength alpha and the fractions of cross-and double cross-slips generation at a cavity having a diameter of 2 nm in pure tungsten were determined as 0.50 +/- 0.21 and approximately 37 % at 973 K, respectively. The fractions of cross-and double cross-slips generation of screw-type dislocations at a cavity were nearly identical at 723 K and 973 K. The types of generated double cross-slips were changed with increase in temperature.
By coupling the exceptional mechanical and durability merits with radiation shielding characteristics, radiation shielding ultra-high-performance concrete (RS-UHPC) promises to foster nuclear safety and security. In this study, a novel RS-UHPC was developed by leveraging the synergistic effects of Ilmenite and Ferroboron as replacements for Quartz Sand. Particle packing was optimized using the Modified Andersen & Andreasen model (MAA), while a simplex centroid design was implemented to capture the separate and joint effects of Quartz Sand, Ilmenite, and Ferroboron on the physical, mechanical, and radiation shielding properties of non-fibrous RS-UHPC. A total of 13 mixture designs, 9 for model establishment and 4 for verification, were developed. Density and compressive strength were experimentally determined, while radiation shielding parameters, including linear attenuation coefficient (& micro;), mass attenuation coefficient (& micro;m), thermal neutron capture crosssection (& sum;abs), and effective removal macroscopic cross-section (& sum;R), were analytically evaluated. The developed RS-UHPC mixtures achieved densities up to 3,196 kg/m3 and compressive strengths up to 125 MPa, while using a deliberately selected higher water-to-binder ratio that, despite slightly reducing mechanical properties, it significantly enhanced neutron attenuation efficiency. Model adequacy was verified, and mixture optimization based on a pre-defined desirability function tailored for nuclear applications identified an optimal ternary system of 28% Quartz Sand, 22% Ilmenite, and 50% Ferroboron. Compared to the reference mixture, the optimized mixture improved the density, & micro;, and & sum;R by 31, 28, and 24%, respectively, while achieving a remarkable enhancement in & sum;abs exceeding 1,600%, owing to the high boron content. These findings contribute to propel the discovery of advanced shielding materials needed for fostering nuclear safety and security.
