Journal of Nuclear Materials最新文献

Master curve (MC) testing of VVER-440 RPV surveillance specimens treated for 27 years (∼200,000 h) in a surveillance channel of Metsamor-NPP was performed to investigate the influence of long term high fluence irradiation on RPV embrittlement. The surveillance chain consisted of both thermal aged specimens (above the core level) and irradiated specimens (inside the core). The reference temperature (T_o) values obtained from irradiated specimens are compared with the results from thermal aged specimens to characterize irradiation induced shifts in T_o values for both base and weld metal specimens. It was found that the high fluence irradiation up to a nominal fluence of 3.2 × 10²⁵ n.m^-2 at E > 0.5 MeV resulted large embrittlement with T_o shift values greater than 300 °C for both base and weld metal specimens. The obtained shifts in T_o values at these high fluence values were used to compare with the predictions from PNAE procedure in the Russian regulatory guide outside its validity range. It was found that the measured shift in reference temperature for the weld metal was well below the predicted value while, the shift in reference temperature for base metal was largely under-predicted at these high fluence values.

In this work, we investigated the impact of different concentrations of helium (He) impurities on the electronic thermal transport properties of tungsten plasma-facing materials (W-PFMs) at finite temperatures using the W-He tight-binding (TB) potential model. We found that the electronic transport performance decreases with increasing He atom concentration at different sites, where the greatest reduction in the electrical conductivity of the system is caused by the introduction of He atoms at neighboring tetrahedral sites. As the temperature increases, the electrical conductivity decreases, while the electronic thermal conductivity increases. Importantly, the higher the temperature is, the weaker the response of the electrical conductivity and electronic thermal conductivity to the He atom concentration. We suggest that this behavior is attributed to the diverse contributions of scattering mechanisms within various temperature ranges. Furthermore, as the temperature increases, the electron scattering mechanism gradually transitions from electron-impurity scattering to electron-electron scattering. Additionally, our calculated atomic resolved electrical conductivity data indicate that at lower temperatures, the electrical conductivity is predominantly contributed by W atoms around the He cluster.

Based on a critical analysis of the existing model for the effect of excess vacancies and interstitials on the nucleation of coherent particles under irradiation, a new nucleation model is developed that uses the Brailsford-Bullough model for the steady state occupation probabilities of vacancies and interstitials at the particle interface, recently refined by the author. It is shown that, depending on the surface tension $\gamma$ of a coherent particle, the nucleation mechanism can be qualitatively different. In the case of a relatively small $\gamma$, an instability can arise in the irradiated solid solution leading to the barrier-free nucleation of coherent precipitates. In the opposite case of a relatively large $\gamma$, the classical one-dimensional theory of homogeneous nucleation is applicable. In order to eliminate the uncertainty in the magnitude of surface tension (from the literature) and to better understand the underlying mechanisms of coherent particle nucleation in irradiation tests, additional atomistic studies are recommended.

The mechanical behaviors and deformation mechanisms of Chinese low activation martensitic (CLAM) steel under extreme loading conditions were systematically studied. The mechanical experiments were performed at a wide range of strain rate (from 0.001 to 3500 s^-1) and temperature (from 300 to 1073 K). The main results show that the strength of the CLAM steel shows an apparent positive strain rate and temperature softening effect. In particular, at quasi-static loading conditions, the elongation of CLAM steel first decreases (300–673 K) and then increases (673–1073 K) with the temperature rising. Under dynamic conditions, the elongation of the CLAM steel is positively correlated with temperature rising and is larger than that under quasi-static loading conditions. The microstructure characterization results indicate that grain refinement during deformation and the positive strain rate effect on elongation are primarily governed by changes in grain size, especially at high temperatures. The relationship between the plasticity capability, precipitates and grain refinement are also analyzed. The obvious competitive mechanisms under different loading conditions in the recrystallization process of the CLAM steel. In summary, precipitates contribute to grain refinement in martensitic structures by providing nucleation sites for new grains and by obstructing dislocation movement, thereby raising the local strain and promoting dynamic recrystallization (DRX). Both of these mechanisms result in a finer and more uniform grain structure, which enhances the mechanical properties of the material, such as strength and toughness.

To prepare for the future deep geological disposal of high-level waste, the alteration of two inactive nuclear glasses was studied in the presence of claystone, iron, and cementitious grout at 50 °C for a period of 4 years. The materials were immersed together in a synthetic water representative of claystone solution composition. Blank experiments were also carried out to study the independent behaviors of iron and glass in the synthetic water. The evolution of the solutions’ chemistry was monitored. At the end of the experiments, the glass and iron, as well as the neoformed phases, were characterized. Results showed that Si, Ca, Mg, and Fe are key elements. A competition exists between the retention/integration of species into glass gels and clay-like neoformations. Magnesium tends to be leached from glasses and form Mg-silicates. It would only integrate into the gel if it was widely available in the solution. Furthermore, iron could form Fe-silicates. Whether for Mg or Fe, the formation of silicates was detrimental to the glass since it involved the creation of a silicon sink and thus the conservation of a high alteration rate. Concerning calcium, in these experiments it appeared that this species tends to be integrated/retained within the gels. In the pH conditions existing here, it seems that there is no competition with a neoformation. The specific impact of the cementitious grout was also studied, and the results showed that the presence of this material in the system had a beneficial effect on glass alteration due to a significant Si supply in the solution, enabling a reduction in the glass alteration rate.

Irradiation induced growth is a constant volume shape change that occurs without externally applied stress that is observed in some materials under irradiation damage. Proton irradiation was carried out on a pure Zr sample to many dpa, to enable investigation of microscale aspects of the irradiation growth phenomenon. Irradiation induced a significant surface morphology change in the irradiated area, which is believed to be the result of the anisotropic growth behavior of the hcp structured Zr. The localized strain that developed was characterized by Electron back scatter diffraction (EBSD), on both the irradiated surface and on a cross sectional through-thickness plane. It was found that there is a correlation between the amount of local deformation and level of misorientation existing between two adjacent grains. The irradiation induced defect microstructure was characterized by transmission electron microscopy (TEM), showing 〈a〉 and 〈c〉 component loops similar to that generated by neutron irradiation in literature. Lastly, site specific focused ion beam (FIB) TEM lift-outs were prepared on local grain boundaries to investigate the origin of the localised deformation.

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 10¹⁹ n/cm², and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH₂), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Liquid lead (Pb) is a promising coolant for lead-cooled fast reactors, and has advantages over lead–bismuth eutectic (LBE) in many aspects. But it can also cause severe corrosion to structural materials at high temperatures. In this study, the chemical states of 10 alloying elements as well as H and O in liquid Pb were investigated via ab initio molecular dynamics, and the local structure, dissolution and charge transfer were analyzed systematically. In liquid Pb, the coordination numbers (CNs) of Fe and Ni are the smallest, with the value of 7.72 and 7.01, respectively, and Al, Mn, Mo, Nb, Ti, and Si have the biggest CNs with about 12. This is correlated with the effective atomic radii in liquid Pb. Judged by the dissolution energy, the tendency of alloying elements to dissolve in liquid Pb is Al > Si > Ni > Ti > Cu > Fe > Mn > Nb > Cr > Mo. The calculated Bader charges show that the alloying atoms undergo minimal electron transfer in liquid Pb except Ti. H₂, H₂O, and O₂ will decompose quickly and cannot exist in liquid Pb. Then, the alloying element will form a strong binding with O atom, especially Al and Si. O atoms can lead to an increase in the coordination of alloying atoms in liquid Pb and obviously reduce their diffusion coefficients in most cases. These findings can provide guidance for predicting material corrosion and designing future high-performance corrosion-resistant materials.

Recent research indicates that one of the leading candidates for Accident Tolerant Fuels (ATF) are chromium-based coatings deposited on the commercially used zirconium alloys. The chromium-based coatings seem to improve the corrosion kinetics of underlying zirconium in both primary water and steam, where zirconium fails to meet the safety requirements. It is well known that the corrosion kinetics of zirconium can be negatively influenced by high concentrations of lithium ions. To evaluate the effect of chromium-based coatings on corrosion behavior of zirconium alloys in water containing elevated concentrations of lithium, Cr and CrN/Cr multilayer coated Zr1Nb alloy was tested in high temperature water containing 70 ppm Li⁺. Using Electrochemical Impedance Spectroscopy (EIS), Scanning Electron Microscopy (SEM) and Energy-dispersive X-ray Spectroscopy (EDX), we have found that the chromium coating could have a positive effect on the underlying material as it behaves as a good diffusion barrier for oxygen and Li⁺ions. On the other hand, CrN/Cr coating, due to not sufficient structural integrity of the multilayer, showed non-protective behavior to the Zr1Nb alloy. Our results suggest that certain chromium coatings can significantly enhance corrosion resistance in lithium containing water environments.