Journal of Nuclear Materials最新文献

This study examines the microstructural characteristics of the primary water stress corrosion cracking (PWSCC) crack tip in the transition zone (TZ) of the 52 M overlay and its implications for intergranular cracking. In the TZ, the Cr content at grain boundaries near the fusion boundary (FB) is lower than those farther from the FB. Additionally, residual strain at grain boundaries near the FB is higher than that farther away, with the fine-grained zone around the FB showing higher local deformation than the surrounding columnar grains. Stress corrosion cracking (SCC) crack growth rate (CGR) test results indicate that the TZ showed certain SCC sensitivity, and the closer to the FB, the lower the SCC resistance. Analysis of the SCC crack tip found that the distinctive composition distribution of the 52 M overlay TZ, characterized by low Cr and high Fe near the FB, is prone to intergranular oxidation, thereby reducing SCC resistance. Conversely, higher Cr and lower Fe content at grain boundaries in the TZ farther from the FB form dense, Cr-rich oxides ahead of the crack tip that slow SCC crack growth, resulting in the diffusion of oxidation along the dislocation structure into the grains and forming a fibrous oxidation zone.

A composite UN fuel containing 10wt% UB₂ has been manufactured via spark plasma sintering using different milling methods prior to sintering, and the resulting pellets characterised to understand the effects of UB₂ location and morphology on UN sintering behaviour and oxidation performance. Differences in microstructure and phases present were observed, with planetary ball milling leading to smaller UB₂ inclusions as well as the formation of a UBN phase on sintering. Composite pellets showed an increase in the steam oxidation onset temperature when compared to UN at similar density and manufactured from the same feedstock. Of particular note was the behaviour of one sample with a comparably low density (∼92 %) which had an onset temperature of 823 K and a significantly reduced rate of reaction compared to monolithic UN at similar density. This provides the first confirmatory evidence that UB₂ limits the UN-steam reaction by some other mechanism than simply promoting a high-density microstructure. This is supported by examination of post-oxidation composite material, which shows a varied and more complex morphology compared to reference UN samples, including large apparently-bound agglomerates and limited free fine particulate.

Uranium dioxide (UO₂) is the standard fuel used in light water reactors (LWRs). However, it has a low thermal conductivity that ultimately limits its performance both during normal operation and in accident conditions. Adding a material with high thermal conductivity is a potential approach to enhance the thermal conductivity of UO_2. Forming an interconnected structure of high-conductivity material can significantly enhance the overall thermal conductivity of the composite. Molybdenum (Mo) has been used as an additive material in UO₂ composites previously. A new method for the fabrication of interconnected UO₂₋Mo composites using pre-sintered UO₂ granules to improve the continuity of Mo channels was investigated in this study. UO₂–10 wt% Mo composites were produced using UO₂ granules and 1073 K and 1473 K pre-sintered UO₂ granules, followed by spark plasma sintering (SPS) of the mixtures at 1473 K for 5 min. The composites were characterised using scanning electron microscopy and X-ray diffractometry and their thermal conductivities were measured by the laser flash method and compared with a reference UO₂ pellet. At a maximum measurement temperature of 1073 K, a 52 % increase in thermal conductivity was observed in the composites containing UO₂ without pre-sintering, and UO₂ pre-sintered at 1073 K. The increase was 31 % for composites manufactured from UO₂ pre-sintered at 1473 K. These results suggest that higher temperature pre-sintering may be detrimental to forming interconnected Mo structures.

The kinetic coupling between vacancy flux and solute flux is crucial for comprehending and forecasting the microstructural evolution of tungsten-based alloys under irradiation. We utilized the Atomic Kinetic Monte Carlo (AKMC) method to systematically explore the vacancy-mediated diffusion of transition metal (TM) solutes in tungsten. The diffusion and transport coefficients of TM solutes were obtained and then used to identify the occurrence of vacancy drag and solute-segregation tendencies. Our findings indicate that TM solutes exhibit faster diffusion rates than tungsten self-diffusion. Ti, V, Nb, Mo, and Ta do not experience vacancy drag, whereas for other TM solutes, vacancy drag is the primary vacancy-mediated diffusion mechanism at low temperatures, transitioning to the inverse Kirkendall mechanism at high temperatures. Moreover, their transition temperatures were determined, showing a parabolic trend in each TM series with peaks observed at Co, Rh, and Pt for the 3d, 4d, and 5d series, respectively. In the temperature range investigated here (600∼3000 K), Ti, Zr, Hf, V, Nb, Ta, and Mo exhibit depletion at vacancy sinks, while other TM solutes enrich at vacancy sinks due to vacancy drag at low temperatures but deplete because of the inverse Kirkendall effect as temperature increases. Additionally, our AKMC results confirmed that, for the solute-vacancy interactions, purely attractive and repulsive interactions lead to the vacancy drag and inverse Kirkendall effect, respectively. While for the complex case, involving both attraction and repulsion, the first-nearest-neighbor (1nn) attraction plays a crucial role in enabling solute diffusion via vacancy drag. Even with a strong 2nn repulsion, the solute can still diffuse through vacancy drag, provided there is a 1nn attraction.

The Lead-Bismuth Fast Reactor (LBR) emerges as a promising concept due to its superior neutron economy, chemical stability, and thermal properties. From the nuclear safety standpoint, the focus has predominantly been on the behavior of ²¹⁰Po and other fission products, while the issue of tritium in LBRs has not been sufficiently addressed due to the inconvenience of tritium experiments. Formation of tritium/helium bubbles induces significant local stress and volume expansion, leading to hardening and embrittlement of structural materials, thus expediting their degradation through irradiation effects. Current understanding of tritium transport within liquid Lead-Bismuth Eutectic (LBE) remains incomplete. To bridge this gap, a novel device employing the "permeation pot" method has been developed for the first experimental quantification of diffusivity, permeability, and solubility of hydrogen isotopes in liquid LBE. Notably, hydrogen diffusivity in this medium is found to be three to four orders of magnitude greater than in conventional 316L stainless steel structural material. Furthermore, the temperature-dependence of diffusivity in liquid metals is minimal compared to solids, as indicated by the activation energy. Conversely, solubility in 316L significantly surpasses that in LBE by three to four orders of magnitude. This discrepancy accelerates the tritium release from LBE to structural material, leading to the failure of the structural material.

The high-temperature oxidation performance of Cr-coated Zr-4 alloy in air and steam atmosphere is comparatively studied and the mechanism of steam promoted oxidation is revealed by density functional theory (DFT) calculation. The experimental results show that there are significant differences in surface and cross-sectional microstructures after oxidation in the two atmospheres. Dense irregular polyhedral oxides accompanied by randomly occurring micro-cracks are developed after high-temperature air oxidation. While mackerel scale-like or worm-like particles with whisker-like structures accompanied by defects such as pores and micro-cracks are developed after high-temperature steam oxidation. In the high-temperature steam atmosphere, the more vigorous atomic diffusion leads to a thicker Cr-Zr diffusion layer and higher O content, so that after exposure at 1100 °C for 3 and 4 h, the Zr-4 alloy adjacent to the Cr-Zr diffusion layer is oxidized to ZrO₂. All the experimental results demonstrate that Cr-coated Zr-4 alloy experiences more severe oxidation in high-temperature steam atmosphere. The DFT calculation results reveal the main reason of steam promoted oxidation is that the interstitial H protons boost the formation of Cr and O vacancies and vacancy pairs in the Cr₂O₃ oxide scale.

Simulated pressurized water reactor conditions (330 °C, 15.6 MPa, ∼20 ppb oxygen) without irradiation were used to investigate the hydrothermal corrosion behavior of ultrasonic additively manufactured Zircaloy-4 up to 1000 h. X-ray computed tomography allowed for visualization of defects from processing and their progression after corrosion experiments. The specimens were found to have clear variability in the mass change data, compared to typical wrought Zircaloy-4 specimens. The variation in the mass change after exposure was attributed to weld defects connected to the specimen surface which allowed ingress of oxidant into the samples. Defects visualized by computed tomography were found via metallography and characterized. Ultrasonic additively manufactured Zircaloy-4 was found to have comparable corrosion behavior as wrought Zircaloy-4 for specimens which did not have clear surface defects along weld interfaces.

Interest in controlled deuterium-tritium fusion as a clean-energy technology has grown in recent years. Solid tritium breeder materials, such as γ-LiAlO₂, need to release tritium to allow the fusion reaction to occur and control over the microstructure can help tune tritium release. Internal gelation is a synthesis technique that allows control over the sample's microstructure. In this process, droplets of an aqueous precursor are heated to form a gel, which is washed, dried, and calcined to produce oxide spheres. Past attempts applying internal gelation to fabricate γ-LiAlO₂ revealed that lithium was lost during wash processes, leading to a lithium deficient final product. To overcome lithium deficiency, the chemistry and reaction pathway must be unraveled. Therefore, this work strove to elucidate the mechanisms of lithium aluminate formation and lithium deficiency. Complementary characterization techniques revealed that lithium aluminate produced via internal gelation formed an aluminum-lithium layered double hydroxide (LDH) structure as an intermediate species. This LDH has a stoichiometry of Li(Al(OH)₃)₂NO₃·xH₂O and the 1:2 ratio of Li to Al is thought to limit the overall lithium content of the final lithium aluminate product due to loss of unbound lithium during washing. This work also indicates that infusing amorphous aluminum hydroxide samples with lithium after gelation appears to incorporate lithium beyond this LDH stoichiometric limit and is one potential pathway towards the desired γ-LiAlO₂ final product. The results of this work highlight the influence that the LDH intermediate species has upon the formation and stoichiometry of the final, calcined product.