In this work, metal-ceramic carbides exposed to eutectic lithium-lead have been tested in the COrrosion Experiment in Static Conditions facility. A first investigation on chemical compatibility with silicon-infiltrated silicon carbide was undertaken in a lithium lead environment at 550°C for 1000 hours with an argon-inert atmosphere, representative of fusion breeding blanket operating conditions. This experiment involved the prior implantation of helium and hydrogen in order to simulate the potential influence of light ions resulting from neutron transmutation on the matrix properties. The findings demonstrated good compatibility for all specimens, confirming the wide applicability of silicon carbide materials to breeder blanket purposes. A second compatibility test was conducted under identical temperature and duration conditions as the previous experiment, but in an argon gas flow with partial O₂ pressure to assess the oxidizing effect. The samples subjected to this treatment included chromium carbide, tungsten carbide, and silicon carbide. Microscopic and spectrometric investigations revealed a degraded layer for Cr3C2 and WC samples, whereas the Si/SiC matrix remained unaltered regarding chemical and structural composition, evidencing its high resistance under these adverse conditions.
Despite unavoidable, the effect of oxygen contamination on the thermophysical and structural properties of molten salts remains largely unexplored. To elucidate how oxygen influences chloride-based molten salts with uranium, we analyze two distinct scenarios where oxygen is present either as molecular oxygen (O2) or oxide ions (O2-). Specifically, molecular dynamics simulations based on universal machine learning interatomic potentials (MLIP) are performed on KCl–UCl3, NaCl–UCl3, and NaCl–KCl–UCl3 melts at 1,000 K with oxygen contents ranging from 0 to 2.0 wt%. The simulations show that O2 and O2- modify the polyhedral network of chloride melts in distinct ways, leading to contrasting trends in density. In the O2 gas ingress scenario, the density decreases, primarily driven by compositional dilution from the addition of light oxygen species. Structurally, strong U–O bonding induces localized densification that partially mitigates this reduction. However, the emergence of rigid, near-linear U–O–U motifs connected via corner-sharing imposes geometric constraints that lower packing efficiency, further contributing to the density decrease. Crucially, this structural reorganization displaces tightly bound chloride ions to active sites, thereby providing a thermodynamic driving force for corrosion. In contrast, O2- ion accumulation effectively replaces chloride ions, and triggers a global contraction of the anion framework. This contracted environment subsequently facilitates the formation of dense, high-connectivity uranium clusters connected via polyhedral edge- and face-sharing, synergistically driving a consistent increase in density. Our analysis indicates that O2 gas ingress and O2- ion accumulation have separable effects on density, arising from fundamentally different structural responses and packing mechanisms.
Chromium (Cr) is a potential coating material for mitigating fuel-cladding chemical interaction (FCCI) in advanced nuclear reactors. However, its response to high-dose irradiation remains insufficiently understood. In this study, the irradiation behavior of Cr-coated HT9 ferritic/martensitic (F/M) steel was investigated under 3 MeV Fe-ion irradiation to a peak dose of 74.5 displacements per atom (dpa) at 500°C for 61 h. The results show that the void swelling ranges from 0.14% to 0.17% at depths of 100–300 nm under 33.6 dpa, while it decreases to 0.04% near the Cr/HT9 interface at a depth of approximately 500 nm under 46.7 dpa. The columnar grain structure of the magnetron-sputtered Cr coating suppresses swelling along grain boundaries. No obvious cracks or significant elemental diffusion are observed at the Cr/HT9 interface, and within 100 nm on each side of the interface, the swelling is 0.04% in Cr and 0.02% in HT9. These findings demonstrate good interfacial stability under irradiation, suggesting that the Cr coating exhibits promising potential as a barrier against FCCI under high-dose irradiation.
Fission gas release modelling is a key aspect of fuel performance assessment, and quantifying the associated uncertainties remains a challenge. This work performs two sensitivity analyses to build confidence in recently developed athermal and burst fission gas release models. The computational framework adopted for fuel performance simulations couples the fission gas behaviour module of SCIANTIX with the TRANSURANUS fuel performance code. This system is interfaced with Dakota software to perform a global sensitivity analysis based on variance decomposition. Five parameters are investigated for the athermal release in the US PWR 16x16 base-irradiation case, i.e.: fabrication porosity, grain-edge length, burn-up, temperature, and fission rate. Results highlight the dominant role of fabrication porosity, suggesting the need for future modelling efforts on fuel densification. For the burst gas release from fuel microcracks five parameters are studied in the HATAC C2 rod irradiation experiment, i.e.: grain-boundary energy, bubble radius and internal pressure, temperature, and hydrostatic stress. Grain-boundary energy, which controls the microcracking threshold, emerges as the most relevant parameter. Furthermore, the analyses highlight the need for enhanced lower-scale physical modelling to further reduce empirical assumptions. The outcome of this study is twofold: an operational range of parameter values resulting in a calculated fission gas release within a factor two from the experimental value, and a methodology for testing and interpreting model behaviour within the TRANSURANUS–SCIANTIX framework prior to engineering-scale validation.
Tungsten (W), widely to be used as a fusion divertor material, suffers from irradiation-induced embrittlement caused by displacement damage and transmutation helium (He). In this study, micro-cantilever specimens were fabricated in recrystallized pure W by focused ion beam milling and bent by nanoindentation to separately quantify grain-interior and grain-boundary strengths under three conditions: unirradiated, neutron-irradiated (0.74 dpa at 830 °C), and He-implanted W (200 appm, 50 MeV He2+) followed by post-implantation annealing (PIA; 1500 °C for 1 h). Grain-interior micro-cantilever specimens exhibited ductile bending without fracture, accompanied by slip-band formation. The mean maximum grain-interior bending stresses for unirradiated, neutron-irradiated, and He-implanted W plus PIA were 920 MPa, 1092 MPa, and 1454 MPa, respectively. In contrast, all grain-boundary micro-cantilever specimens exhibited brittle fracture along grain boundaries. The mean maximum grain-boundary bending stresses were 872 MPa for unirradiated W and 839 MPa for neutron-irradiated W, indicating no pronounced grain-boundary weakening under the present neutron-irradiation conditions. For He-implanted W plus PIA, grain-boundary stress exhibited large scatter, with high-strength boundaries (mean 1167 MPa) and weak boundaries (mean 417 MPa). Some grain-boundary cantilever specimens fractured at boundaries not aligned with the loading direction, suggesting preferential failure along weak sub-grain boundaries. These results support a Ludwig–Davidenkov (L–D) interpretation in which neutron irradiation is dominated by hardening embrittlement, whereas He promotes non-hardening, grain-boundary embrittlement at specific boundaries.
Understanding the atomistic origin of the initial hydrogenation of yttrium (Y) is essential for rationalizing the nucleation of the δ-YH2 phase and the evolution of α-Y/δ-YH2 interphase boundaries. Here, density functional theory calculations are performed to study early-stage hydrogen incorporation on three low Miller-index facets (0001), (101¯0) and (112¯0) of α-Y, and hydrogen transport across the corresponding α-Y/δ-YH2 interfaces. Hydrogen adsorption is exothermic on all three facets, indicating a thermodynamic driving force for surface hydrogenation. With increasing hydrogen coverage, the Y (0001) and Y (101¯0) facets undergo spontaneous interfacial slip, yielding interfacial geometries consistent with YH2 (111) and YH2 (110) orientation relationships, respectively, which suggests a crystallographic pathway for hydride nucleation at the advancing front. For Y(0001)/YH2(111) interface, an interstitial H atom relaxes spontaneously toward the tetrahedral site. In contrast, H migration toward tetrahedral sites across the Y(112¯0)/YH2(100) and Y(101¯0)/YH2(110) interfaces requires overcoming diffusion barriers of 0.52 and 0.46 eV, respectively. These results suggest that early-stage hydrogenation of Y is promoted by the interplay of hydrogen diffusion and slip-assisted structural rearrangements at the α-Y/δ-YH2 interface.
Cold rolled pure tungsten (W) and sintered W-3Re (Rhenium) alloy, used as a surrogate for selected compositional and microstructural characteristics expected during long-term operation of W plasma facing materials (PFMs) in DEMO-like fusion reactors, were exposed to simulated plasma transients. Two ion flux ratios were used, pure He+ and reactor relevant He+:D+ ion ratios of 10:90, while all experiments were conducted at a steady state temperature of 1273 K with transient heat loadings applied via a pulsed millisecond laser. Pre-irradiation characterization via electron backscatter diffraction (EBSD) confirmed an average grain size of ∼3-5 µm, while a significantly reduced hardness was observed for W-3Re (∼315-335 HV) compared to W (∼510-560 HV). He+ only exposures lead to the formation of small pores on both materials accompanied by rough shale-like features observed primarily on pure W. In contrast, mixed ion exposures produced smoother surface morphology while simultaneously increasing the appearance of net erosion, measured using Mo witness plates. Added D+ reduced pore size and suppressed surface roughening associated with He+ irradiation, while during simulated ELM-like heat load tests, the pores increased in size with increasing transient power. Splash-like features on the material surface were observed for the highest transient power load during transient only and synergistic He+ conditions but absent in mixed ion tests. W-3Re samples exhibited severe cracking, which is hypothesized to result from the reduced mechanical and thermal properties, as well as differences in material processing. Cracking in W-3Re was more dispersed under mixed ion loading, reducing surface roughness, which may indicate sub-surface gas accumulation influencing stress relaxation. These findings highlight the need for further testing of PFMs under relevant reactor conditions. The surrogate W-3Re samples overall showed more detrimental damage, thus motivating this and future research, specifically related to W PFMs transmutation and thermal annealing properties.
Irradiation creep can alter core geometry and potentially relax torque in screws or mechanical springs, thereby affecting overall system performance. A major contributor to the increased creep rate is enhanced dislocation multiplication and migration. Traditionally, irradiation-driven dislocation multiplication has been attributed either to irradiation-induced point defects that enable dislocation climb through the Bardeen–Herring mechanism, or to interactions between irradiation-induced dislocation loops that generate mobile dipoles. In this study, molecular dynamics simulations of body-centered cubic (BCC) tungsten reveal a distinct mechanism of cascade-enhanced Frank–Read dislocation multiplication. Specially, the coalescence of irradiation induced loops establishes an edge-dominated dislocation network that form the structural basis for multiplication, whereas cascade-induced transient stresses facilitate the glide and bowing of pinned edge dislocations within this network, thereby promoting dislocation multiplication and contributing to the increase in creep strain. Since these cascade-induced transient stresses occur only during irradiation, the present results suggest that cascade-enhanced dislocation multiplication may be one of the factors contributing to the higher creep rates observed under irradiation.
We describe an attempt to build a multiscale model for the prediction of the behaviour of point defects in uranium mononitride based on ab initio electronic structure calculations. We discuss in detail the construction and tuning of machine learning interatomic interaction models for uranium mononitride. The calculation of point defect formation free energies and their concentrations at different conditions is considered together with defect migration energies. The resulting model of point defects allows us to calculate the stability boundary for hypo-stoichiometric uranium mononitride and to estimate the self-diffusion coefficients of U and N. We compare our results with the experimental data and discuss the limitations of our model, highlighting the problems that, as we see, remain open.
Severe tempering deformation was successfully introduced into the reduced activation ferritic/martensitic (RAFM) steel, yielding fine fibrous grains along the rolling direction as well as dense and uniform dispersion of M23C6 nanoparticles along subgrain/grain boundaries. Compared to conventional RAFM steel, this tempformed steel exhibits a 35% increase in yield strength and a 15.1% increase in total elongation at 600°C due to its superior microstructural thermal stability. More weak interfaces of the tempformed steel such as those in fibrous grains and at the matrix/M23C6 boundaries facilitate crack delamination and branching, thereby imparting exceptional low-temperature toughness-typically exceeding 179.0±7.4 J/cm2 at -80°C. However, the tempformed steels exhibit a slight decrease in creep resistance and pronounced mechanical anisotropy.
