Journal of Nuclear Materials最新文献

The phase transition behavior of RAFM steel during Hot Isostatic Pressing (HIP) diffusion bonding process would extensively impact the service performance of blanket component in the future fusion reactor. In this paper, the influence of cooling behavior on phase transformation and mechanical property of RAFM steel was studied using a combination of Electron Backscatter Diffraction (EBSD) and a thermal dilatometer. The results show that as the cooling rate decreases after austenitizing, the phase category transitions from a single-phase of martensite to the dual-phase of martensite and ferrite, along with an increase in the rate of ferrite, minor martensite packets and high-angle grain boundaries (HAGB). As the cooling rate decreases, the initial temperature for martensite transformation significantly increases, and the transformation driving force slowdown. Moreover, specimens processed with HIP cooling show lower strength, higher elongation, and a greater standard deviation of hardness compared to air cooling. These differences can be attributed to the formation of ferrite and the diffusion of carbon atoms within martensite. The research findings reveal the phase transition and microstructural evolution of martensite under varying cooling conditions, providing a reference for the controlling the cooling process after HIP bonding of blanket steel components in the future fusion reactor.

Mechanistic multi-scale modelling holds the potential to inform fuel performance codes by incorporating high-fidelity models, algorithms, parameters, and material properties. In this context, meso-scale codes emerge as valuable tools for developing detailed models and performing separate verification and validation steps. This work focuses on SCIANTIX, an open-source 0D meso-scale code designed to describe the behaviour of gaseous and volatile fission products in nuclear oxide fuel. The code predominantly employs engineering physics-based behavioural models featuring computational times that align with typical fuel performance code requirements. Given the numerical foundation of the code, it is applicable to both stationary and transient conditions. Following a recent work outlining the standalone SCIANTIX (version 2.0) performance and its separate-effect validation database, we present its performance when coupled with fuel performance codes to simulate light water reactor fuel rods. The experiments selected for the comparative analysis constitute an initial integral validation database. The comparison focuses on conventional engineering quantities of interest, such as integral fission gas release, demonstrating the satisfactory performance of the code. Additionally, it highlights the potential advantages of multi-scale modelling over conventional semi-empirical approaches.

Zirconium (Zr) alloys are widely used as structural materials in the core of nuclear reactors. However, the absorption of hydrogen by these alloys during the reactor operation may lead to the precipitation of hydrides. Such hydrides have considerable effects on the microstructure and mechanical properties of the alloys, thus significantly impacting their overall performance. In this work, we investigate the behavior of hydrogen in the presence of symmetric tilt grain boundaries in Zr using molecular dynamics simulations. To systematically explore the conditions for the formation of hydrides, we consider hydrogen concentration levels of up to 50 at.%. We find that hydrogen concentrations of 30 at.% or lower are below the threshold for the precipitation of hydrides, while fcc coordinated hydrides precipitate at 50 at.%, which is very close to the solubility limit of hydrogen in Zr at 800 K. We find that the grain boundaries (GBs) play a pivotal role in the behavior of hydrogen, as well as the formation of hydrides in Zr systems. In most cases, large amounts of hydrogen are found to accumulate at GBs. However, both homogeneous and heterogeneous nucleation are observed during the formation of hydrides at the grain interiors and GBs, respectively. Nonetheless, in some cases the GB inhibit the formation of hydrides.

The microstructure and element distribution in the transition zone (TZ) of Alloy 52M buttering (52Mb) layer near the fusion boundary to the SA508III low alloy steel in a dissimilar metal weld joint with post-weld heat treatment (PWHT) is characterized. Two macroscopic TZs in the heat-affected zone have lower Cr and Ni contents than the bulk Alloy 52Mb. The first TZ near the fusion boundary has the lowest Cr and Ni compositions, accompanying the highest Fe content. Cr depletion and Ni enrichment along with slight Fe enrichment at the dendrite boundaries in the TZs, are more significant than those in the bulk Alloy 52Mb. The number and the maximum depth of local oxidation penetrations at the alloy-inner oxide interface based on the oxidation tests, the number and the maximum length of locally interdendritic cracks, as well as the average crack growth rate (CGR) in terms of the stress corrosion cracking (SCC) band based on the SCC tests in the first TZ are higher than those in the second TZ. The average SCC CGR is enhanced by macroscopic element dilution, while local oxidation penetration, as well as locally interdendritic SCC, are thought to be enhanced by both the macroscopic dilution of Cr in the TZs and the microscopic Cr depletion at dendrite boundaries.

The CuCrZr alloy combines high thermal conductivity and mechanical strength with stability at high-medium temperatures, making it a promising heat sink material for the EU-DEMO divertor and limiters. Additive Manufacturing (AM) technologies have proven effective in developing complex-shaped components with almost no constraints on geometry and minimal machining and welding requirements. This makes them particularly suitable for the production of heat exchangers featuring complex cooling channels with intricate inner structures.

This study demonstrates the feasibility to obtain dense CuCrZr via Electron Beam Powder Bed Fusion (EB-PBF) with high thermal conductivity and enhanced mechanical strength compared to conventional routes. Gas atomization was used to produce spherical powders with a composition close to ITER specifications. By optimising the EB-PBF process parameters, relative density values of 99.7 % were achieved after HIP treatment, that removes the eventual residual porosity. The results underscore the importance of meticulous powder manufacturing to mitigate oxidation and microstructural defects in the final components. Achieving high relative densities in the EB-PBF process requires a focus on adopting high-energy absorption rates in the powders. This strategy can be accomplished by reducing the scanning speed and consequently the building rate of the process. The microstructural characterization revealed a complex hierarchical microstructure composed of grains and grain boundaries, solidification-enabled cellular-like subgrains elongated along the building direction and an ultra-fine precipitate state (already present in the as-built condition) mainly consisting of Cr-rich nanoprecipitates, although Zr-rich precipitates were also found at the melt pool boundaries. The thermal conductivity, hardness, mechanical strength at room temperature and high-medium temperatures were measured and correlated with the EB-PBF process parameters and the microstructure obtained after HIP treatment. The results indicate that it is possible to obtain CuCrZr with improved mechanical behaviour compared to conventional manufacturing technologies, while maintaining the thermal conductivity requirements for EU-DEMO.

Increasing the average grain size of fuel pellets by doping them with chromium oxide is one strategy to improve oxide nuclear fuels performance. The promoted fission gas retention is thought to improve the performance of the fuel at high burnup. In this work, we review models for the solubility of chromium in UO₂, and the evolution of the chromium phases in the fuel matrix during irradiation. These models are implemented in SCIANTIX, an open-source mesoscale code describing inert gas behaviour in nuclear fuel. We adjusted the chromium solubility model keeping each parameter within its range of compatibility with experimental data, targeting a better representation of available electron probe microanalysis data of chromium content in fuel after irradiation. As for fission gas behaviour, we considered a physics-based description of the chromium impact on the fission gas diffusivity in fuel grains. The expression for the fission gas diffusivity in standard non-doped uranium oxide has been extended by introducing the impact of the concentration of defects introduced by interstitial oxygen excess representing the effect of chromium content in the fuel itself. A preliminary integral assessment of the proposed models has been carried out against the available experimental data.

Poor mechanical strength at high temperature is the key problem to limit the application of FeCrAl alloys as the candidates for the accident tolerant fuel (ATF) cladding in LWRs. Fe-13Cr-5Al (wt %) alloys were strengthened by adding solid solution element Mo and nanosized ZrC particles, and the strengthening mechanism was assessed by microstructure characterizations including TEM and EBSD. Fe-13Cr-5Al-2Mo-1ZrC alloy has the highest ultimate tensile strength (UTS) and an acceptable ductility at each tested temperature (Room temperature, 400 °C or 800 °C). Especially at 800 °C, the UTS of Fe-13Cr-5Al-2Mo-1ZrC alloy is about 124 MPa, which is 125 % and 34.7 % higher than that of raw Fe-13Cr-5Al (55 MPa) and Fe-13Cr-5Al-1ZrC alloys (89 MPa), respectively. Fe-13Cr-5Al-2Mo-1ZrC alloy has the highest hardness of 306.3 HV, which is 12.8 % higher than that of Fe-13Cr-5Al and 3.7 % higher than that of Fe-13Cr-5Al-1ZrC alloys, respectively. Furthermore, Fe-13Cr-5Al-2Mo-1ZrC samples maintain high strength and favorable ductility after annealing at 1000 °C for 20 h indicating their superior thermal stabilities. The excellent mechanical properties and superior thermal stabilities of Fe-13Cr-5Al-2Mo-1ZrC alloy were not only attributed to the dispersion strengthen by nanosized ZrC particles, the solid solution strengthening by Mo elements, but also the grain refinement structure promoted by the synergistic effects of Mo and ZrC additions.

Lithium orthosilicate (Li₄SiO₄) is regarded as a candidate for tritium breeding in fusion reactors. In this study, the Li₄SiO₄-Li₂ZrO₃ biphasic was developed to improve the sinterability, density, and mechanical properties of Li₄SiO₄. The Li₄SiO₄-xLi₂ZrO₃ (x = 0.5, 1) composite powders were prepared using solid-state reaction via in-situ method. Li₆Zr₂O₇ existed in the ceramic powders at low calcination temperatures. When the calcination temperature was increased to 900 °C, Li₆Zr₂O₇ was transformed into Li₂ZrO₃ due to the decomposition of Li₆Zr₂O₇ at high temperatures. The TEM observation confirmed that the powders consisted of Li₄SiO₄ and Li₂ZrO₃. The Li₄SiO₄ and Li₄SiO₄-xLi₂ZrO₃ (x = 0.5, 1) pebbles were fabricated by the sol-gel method. The measurement results showed that the pebbles had a narrow size distribution and fine sphericity. The density of Li₄SiO₄-Li₂ZrO₃ pebbles reached 96.01% of the theoretical density when it was sintered at 1000 °C for 4 h. Compared with the Li₄SiO₄, the grain size of ceramic pebbles was significantly reduced. Owing to decreased grain size, 139.0 N crush load for the pebbles and 92.4 MPa bending strength for the sintered bodies were achieved. Besides, the ceramics with the Li₄SiO₄ to Li₂ZrO₃ ratio of 2: 1 exhibited preferable mechanical properties. Further, to investigate the chemical stability of biphasic ceramics, the structure and mechanical properties were examined under high temperatures and continuous inert gas purging, simulating the working condition of fusion reactors. It was shown that there was almost no change in phase composition and grain size after purging for 60 h at 650 °C. For the mechanical properties, the crush load was decreased initially due to the cracking in the surface region of the ceramic and then increased because the cracking was recovered.

Deformation of proton-irradiated chromium was investigated using micro-pillar compression and electron microscopy. After 2 MeV proton irradiation at 350 °C, four micro-pillars were prepared from a single grain on the polished specimen cross section. Depending on the distance away from the irradiated surface, hardness as a function of local damage level was studied. All pillars developed a narrow deformation band on one set of near-adjacent {110} planes, arising from closely-positioned parallel gliding. The critical resolved shear stress for gliding along 〈111〉/{110} was measured to be 59.6 MPa in unirradiated material beyond the proton range. The critical stress increased by 20 % after 0.5 dpa, and by 58 % after 1 dpa, with saturation of hardening occurring by 0.7 dpa. Post-compression characterization using transmission electron microscopy showed extensive formation of nanometer size voids in a matrix dominated by tangled dislocations. No twinning was observed. The experimental observations are in good agreement with molecular dynamics simulation of pillar compression of chromium, showing dislocation gliding along $〈111〉$/{110} and $〈111〉$/$\{$112}. The continued stability of chromium for LWR application requires extension of the exposure level from 1 dpa to ∼15 dpa expected for typical fuel pin exposure.

To evaluate the radiation tolerance of polycrystalline materials, the damage effects of Fe and W as typical body-centered cubic metals under uniform irradiation are studied by multi-scale models. A guiding descriptor, the absorption bias (the ratio of the absorption abilities of grain boundaries (GBs) to interstitials (I) and vacancies (V)), is proposed to reflect the radiation tolerance of metals with different grain sizes. Low absorption bias promotes defects annihilation through enhancing I-V recombination and optimally tuning its competition with GB absorption. Polycrystalline metals possess high radiation resistant performance with low absorption bias regulated by grain size and temperature. Furthermore, by comprehensively considering the mechanical property, thermal stability, and radiation tolerance described by absorption bias, nano-crystals are recommended for Fe-based structural materials but coarse-grained crystals for W-based plasma-facing materials. This work reevaluates the radiation tolerance of polycrystalline metals, resulting in new strategies for designing structural materials in nuclear devices through manipulating grain sizes.