•High temperature resistance of fine-grained SiC coating layer and large-grained SiC coating layer in TRISO particles is compared and investigated.•Decomposition and recrystallization process of fine-grained SiC coating layer at high temperature in TRISO particles is studied.•A novel formation mechanism of the pores in SiC layers at high temperature is proposed.

The minimum spouting velocity (Ums) of heavy particles in conical spouted bed at high temperatures are critical importance for nuclear fuel coating, which are different from that at room temperature. However, current re-searches on the Ums were concentrated on the Cold Mockup spouted bed. There were few studies about the Ums of heavy particles in conical spouted bed at high temperatures. In this study, the effect of temperature (373-1273 K) on the Ums is systematically investigated. A new Ums correlation with temperature, cone angle, static bed height, particle diameter and particle density is obtained:Ums = 3.30 x 10(8) center dot (dp)(1.63) center dot pp)0.57 center dot (tan gamma)(1.18) center dot (H0)(1.45)center dot 2Dc (T)- (1.44). The equation extends the application of the Ums at high temperatures and can be applied to determine the Ums value well under different temperature. Besides, the influence mechanism of temperature, cone angle, static bed height, particle diameter and particle density on the Ums has been discussed. This correlation is rec-ommended for the determination of the Ums well at high temperatures and thus provides significant reference for producing coated fuel particles of high quality.

BackgroundInert matrix fuel (IMF) can efficiently convert plutonium and long-lived minor actinides used for preventing the proliferation of nuclear weapons and improving spent fuel disposal, hence has been becoming a hot research topic in recent years. The sol-gel method has the advantage of uniform elemental distribution of the products and the wet operation process is less likely to produce radioactive dust, therefore, it has been used to prepare zirconium-based IMF in the research.PurposeThis study aims to prepare a colloidal solution with good dispersive properties and to obtain IMF microspheres with good sphericity, uniform size, and homogeneous elemental distribution.MethodsFirst of all, ThxZr1-xO2 inert matrix fuel was prepared by an external gelation process, and the sol-gel viscosity was used as the main gelation index. Then, the variation tendency of sol viscosity with c(NH4+)/c(NO3-) was investigated for different metal ions concentrations and different temperatures. Finally, the statistical distributions of colloidal particle sizes were obtained for different metal ions and reaction temperatures by laser particle sizing tests, and the X-ray diffraction (XRD) was used to study the structure of IMF after heat treatment at different temperatures.ResultsThe results showed that the complex gelation parameters and properties can be categorized and quantified using gelation field diagrams. In addition, ThxZr1-xO2 IMF kernels with uniform element distribution, good sphericity, and integral appearance were obtained by optimizing the process parameters. Zirconia showed low solubility behavior in the thorium-oxide system, leading to the generation of a biphasic structure.ConclusionsThe results of this study indicate that zirconium-based spherical IMF microspheres with good performance can be prepared by external gelation method.

In this study, U3Si2 pellets were fabricated by spark plasma sintering (SPS) of U3Si2 clinkers generated by a high-temperature solid reaction between U and Si powders. The influences of SPS operating parameters (dwell temperature from 1000 to 1300 degrees C and pressure from 30 to 90 MPa) on the microstructure (porosity, normal and distorted U3Si2 phase, grain size and shape) of U3Si2 pellets were analyzed. The density of U3Si2 pellets enhanced with increasing temperature and pressure and it reached of the 97.9% theoretical density (TD) at 1300 degrees C/60 MPa. The residual stress caused the generation of a distorted U3Si2 phase in the pellets, and low-temperature annealing at 300 degrees C effectively released the residual stress and eliminate the lattice distortion of U3Si2. The U3Si2 grain size increased with increasing dwell temperature and showed an opposite trend with increasing pressure. The grain growth mechanisms of the SPS-sintered U3Si2 pellets involved temperature-dependent sur-face diffusion and pressure-induced grain growth inhibition. The evolution of the U3Si2 grain morphology from equiaxed to columnar grains was due to plastic strain under SPS conditions.

Thorium dioxide as an excellent matrix with stable properties is available for the direct incineration of pluto-nium, while (Th,Pu)O2 is a promising advanced fuel, which can be applied to a wide range of reactor types. In this study, cerium was used to simulate plutonium owing to their similar physicochemical properties. (Th1-xCex) O2 kernels were prepared by an external gelation process, which employs thorium and ceric ammonium nitrate as source materials, ammonium hydroxide as a neutralizing agent, polyvinyl alcohol as a thickening agent, and an ammonia gas ring for preliminary forming. The effect of the reaction temperature, reactant concentration, and pH on the feeding solution gelation properties was investigated. After the dispersive process, the aqueous (Th1-xCex)(OH)y.nH2O gel microspheres were washed and dried at 200 degrees C, calcinated and sintered. It was found that more than 99 % of the sintered microspheres maintained an excellent degree of sphericity and integrality when the parameters were properly controlled. It was found that the feeding solution preparation process was the most important and complex step to obtain uniform microspheres of good sphericity during the whole kernel fabri-cation process. The (Th1-xCex)O2 microspheres produced were characterized using thermogravimetric analysis, X-ray diffraction (XRD), scanning electron microscopy, and energy dispersive spectroscopy (EDS). The XRD analyses showed that all products appeared as a solid solution with a cubic fluorite structure when doped with cerium in the selective compositional range. The elemental scanning by EDS showed an excellent uniform dis-tribution of thorium and cerium in the sintered microspheres.

In this study, a hydrothermal treatment (HT) at 200 °C was used to deal with the cracking of the external gelation-made thorium oxide (ThO2) microspheres and the effects of HT on the composition, microstructure, and cracking fraction of the microspheres were investigated. The results showed that HT removed the light weight impurities such as ammonium nitrate, degraded polyvinyl alcohol with light weight as well as absorbed and bonded water, thus resulting in a reduction of the weight loss of the gel microsphere by 26%. Owing to the removal of the impurities, HT led to the improvement of the crystallinity of the dried microspheres and the increase of specific surface area and pore size of the dried ones. Further, HT had distinct effect on the heat treatment of the microspheres: (1) For the drying process (240 °C) where an atmosphere of humidified air and wet microspheres were essential, the cracking fraction of the treated microspheres was only one-tenth of that of the original ones due to the removal of the bonded water by the HT. (2) In the case of elevated temperature treatment (650 and 1350 °C), the removal of residual polyvinyl alcohol by thermal oxidation was easy for the treated microspheres without obvious occurrence of cracking, however, it was not the case for the original counterparts. With the aid of HT, high-quality crack-free ThO2 microspheres were available after the heat treatment and would be potentially used as nuclear fuel for a solid-fueled thorium molten salt reactor.

A solid-phase metallurgy combined with spark plasma sintering technology was used to prepare U3Si2 pellets. The thermal conductivity and oxidation behavior of the pellets were studied. The pellets were highly dense (> 98

Uranium dioxide (UO2) is currently the most widely used nuclear fuel for commercial nuclear reactors. However, the Fukushima Daiichi nuclear disaster revealed the primary safety risks of this kind of fuel under accident conditions, which thereby stimulated various international programs in developing accident tolerant fuel (ATF). Silicon carbide (SiC) and Zirconium (Zr) are being introduced as additives to enhance thermal properties and fission product retention ability of the UO2 single crystals (SCs) pellet. The oxidation resistance in water at high temperatures is essential to the deployment of ATF. In this work, two kinds of enhanced UO2 single crystal pellets by using SiC and Zr particles as additives (denoted as UO2 SCs-SiC and UO2 SCs-Zr, respectively) were fabricated through spark plasma sintering (SPS) processes. The impact of steam oxidation on UO2 SCs-SiC and UO2 SCs-Zr was investigated through the microstructures, phase compositions and post oxidation performances characterizations. The results revealed that the pellets with SiC as additives showed good oxidation resistance in water even at 1200 degrees C while that with Zr as additives showed complete oxidization which resulted in the formation ZrO2 powder. (C) 2022 Elsevier Ltd. All rights reserved.

BackgroundThe Uranium dioxide (UO2) is currently the most widely used nuclear fuel for commercial nuclear reactors. However, the Fukushima Daiichi nuclear disaster revealed the primary safety risks of this fuel in an accident, so various international programs were launched to develop accident tolerant fuel (ATF), a new generation of fuel system developed to enhance the capability of nuclear fuel assemblies in severe accidents.PurposeThis study aims to improve the thermal conductivity of fuel pellets by adding a second material to the UO2 matrix, an important research direction for ATF.MethodsFirst of all, large-grain UO2 particles were used as raw materials, and the high-density large grain UO2-SiC composite fuel pellets were obtained by Spark Plasma Sintering (SPS) sintering process at lower sintering temperature. Then, the properties, such as microstructure and chemical composition, of the composite fuel pellets were characterized by using metallographic microscope (MM), scanning electron microscope (SEM), X-ray diffraction (XRD) and energy dispersive spectrometer (EDS). Finally, the high-temperature oxidation resistance in air environment was studied.ResultsThe results show that the UO2-SiC interfacial reaction can be avoided by the SPS sintering at lower temperature, and the density of the prepared pellets is more than 95% theoretical density (TD). Compared with traditional UO2 fuel pellets and SPS sintered UO2-SiC pellets using conventional UO2 powders, the thermal conductivity of large-grain composite fuel pellets is significantly improved. Oxidation tests results indicate that the oxidation weight gain of the composite fuel pellets is significantly weaker than that of traditional pellets when the temperature is lower than 350 ℃. However, when the temperature reaches 350 ℃, the oxidation of UO2 cannot be further prevented by SiC.ConclusionsThis study provides reference for improving the thermal conductivity of UO2 matrix by adding a second phase with high thermal conductivity.

Background The solid fuel thorium element molten salt reactors (MSR) have attracted more attention recent years. A3-3 graphite is chosen as the fuel matrix for MSR, thus its irradiation behavior and mechanical property is very important before the application. Purpose The study aims to observe the irradiation defects and hardness of A3-3 matrix graphite after ion irradiation by slow positron beam and nano-indentation, respectively. Methods The matrix graphite of fuel elements was irradiated with 1 MeV Xe ions to fluence of 5.8×1014 ions·cm-2 and 2.9×1015 ions·cm-2 respectively at room temperature. The slow positron beam and nano-indentation were employed to investigate the effect of Xe ions irradiation on vacancy defects and hardness of matrix graphite. The changes in irradiation induced defects distribution with depth and fluence were analyzed according to the obtained positron annihilation S parameters versus positron incidence energy or depth curves, compared to SRIM (Stopping and Range of Ions in Matter) calculation. Results Results from slow positron beam measurement show that 1 MeV Xe ions irradiation in matrix graphite introduces a damage layer with depth of about 600 nm, and the damage peak locates at about 250~350 nm in depth, consisted with SRIM simulation. The S parameters in irradiation samples increase significantly compared to virgin sample, which suggests that a high concentration of vacancy-type defects appeared within irradiation damage layer. In addition, the S parameters increase with the irradiation fluence, which shows that the concentration or size of vacancy-type defects increases. The nano-indentation results show that the hardness of irradiated graphite matrix is enhanced. Conclusions The enhanced hardness of A3-3 matrix graphite after ion irradiation is ascribed to the pinning of basal plane dislocation by the high concentration of vacancy type defects introduced by irradiation, consisted with the slow positron beam analysis. Slow positron beam is a very sensitive tool to study the irradiation defects.

The effect of helium ion irradiation with low and high ion fluences on the corrosion behaviour of nickel-based alloys in FLiNaK molten salt at 700 degrees C was investigated. The microstructure and element distribution were characterized to explore the corrosion behaviour of matrix and grain boundaries. The results showed that irradiation accelerates matrix corrosion in both Inconel 617 and GH3535 alloys. Moreover, the irradiation decelerated and accelerated intergranular corrosion were respectively observed under irradiation with low and high helium ion fluence, which is caused by the combined effects of the self-healing mechanism and diffusion of interstitials and helium atoms.

The irradiation damage behavior was studied in the nano-grained Ni–Mo–Cr alloy (nano-grained GH3535), which was irradiated by He ion to various dose. The evolution of defects and hardness changes are characterized by transmission electron microscopy and nanoindentation to explore the irradiation tolerance of the nano-grained GH3535 and the coarse-grained GH3535 (annealed GH3535), where the later was chosen as reference material to make comparison with nano-grained GH3535. The results show that though both the average size and number density of He bubbles increase with an increase in the irradiation dose, the smaller volume fraction is found in the nano-grained GH3535 compared with the coarse-grained GH3535 under the same irradiation condition. This indicates that the nano-grained GH3535 possess better irradiation swelling resistance than the coarse-grained GH3535. However, the increase in the hardness of the nano-grained GH3535 is more significant than in the coarse-grained GH3535 under the same irradiation dose. This suggests stronger irradiation-induced hardening of the nano-grained alloy comparing to coarse-grained alloy, due to the impeding effect caused by grain boundaries decorated with He bubbles. This study provides insight into the design of irradiation-tolerant nickel-based alloys for nuclear industry applications.

In this study, a novel graphite was prepared by adding MCMBs (mesocarbon micmbeads) to densify matrix graphite A3-3, a typical fuel element and moderator material in high-temperature gas-cooled reactors. The densified graphite was evaluated as a potential fuel element material for molten salt reactors (MSRs). In particular, the effects of MCMB loading on physical properties and the corresponding effects on the infiltration of liquid fluoride salt in MSRs were studied, as molten salt infiltration can impact safe reactor operation. MCMBs with a mean particle size of approximately 3 mu m were added to matrix graphite A3-3 at loading levels of 1 wt%, 5 wt%, 10 wt%, and 15 wt%, and samples were prepared by a quasi-isostatic pressing process at 250 MPa. Mercury porosimetry data showed that MCMB-densified graphite (MDGs) had lower porosity (approximately 15 vol%) and median pore diameters (ranging from 105 to 306 nm) compared to A3-3 itself (22.9 vol% porosity and 573 nm median pore diameter). Importantly, the MDGs exhibited greatly reduced infiltration by FLiBe molten salt compared to unmodified A3-3. For instance, A3-3 modified with 1 wt% MCMB exhibited a 2.1 % weight increase after treatment at 1 MPa for 20 h, while unmodified A3-3 exhibited a weight increase of 11.6 %. This study revealed that MDGs could be promising candidate materials for the graphite matrix of fuel elements as well as a neutron moderator and reflector in MSRs.

Alloy 800H has recently received significant attention owing to its potential applications in molten salt reactors. However, its irradiation tolerance and plastic degradation after irradiation are not well understood. In this study, Alloy 800H was irradiated by 3 MeV Cu+ ions with irradiation damage up to 10 dpa. The microstructure changes were characterized using two Rel-rods cases and weak beam dark field technology by performing transmission electron microscopy. The plastic mechanical properties before and after the irradiation were extracted from P-h curves derived by performing nanoindentation at loading rates of 0.05, 0.1, and 0.5 nm/s. The microscopy studies demonstrate that the characterization and evolution of the two variations of Frank loops are distinct and the 1/3 <(1) over bar 11 > Frank loops firstly nucleate and coarsening. The nanoindentation results demonstrate that the measured yield stress increases with the increasing irradiation damage and decreasing loading rate. The good agreement on yield stress between the calculated and measured values suggests that using the characteristics of two variations of Frank loops to calculate the contribution of yield stress is appropriate. In addition, the decreasing measured values with increasing loading rate are attributed to the increased geometrically necessary dislocations, which are correlated with the indent size effect.

Irradiation damage due to helium embrittlement impacts the safety and integrity of welded structural materials in power reactors. To understand this helium effect, in this study, GH3535 welded joints were irradiated at 650 degrees C with 500 keV He ions at various doses to investigate the effects of He bubbles on the weld, heat-affected zone (HAZ), and base metal. The three regions exhibited different hardening be-haviors. The hardness of the weld and HAZ after irradiation increased proportionately with the ion dose. The hardness of the base metal gradually saturates at a dose of 2 x 10(16) ions/cm(2). At the highest dose (1 x 10(17) ions/cm(2)), the degree of hardening in the three regions was as follows: weld (92.9%) > HAZ (91.3%) > base metal (72.7%). The microstructures of the samples revealed that the mean diameters of the He bubbles in the weld, HAZ, and base metal were approximately 2.74, 2.80, and 2.36 nm, respectively, with corresponding number densities of 15.4 x 10(23) , 10.8 x 10(23), and 9.69 x 10(23) m(-3). Furthermore, the yield strength increment was calculated using the dispersed barrier hardening model, which suggested that the helium bubbles played an important role in the different hardening behavior of the welded joint. The nucleation and growth of helium bubbles was influenced by the structures of the intrinsic disloca-tions in the weld, HAZ, and base metal. (C) 2021 Elsevier B.V. All rights reserved.

To evaluate the effects of irradiation on additive manufacturing materials which are expected to be applied in advanced nuclear energy systems, selective laser melting 316L stainless steel (SLM 316L SS) and traditionally manufactured 316L stainless steel (TM 316L SS), were irradiated with 500 keV He ions at 700 degrees C, respectively. SEM, TEM and nanoindentation measurements were performed to survey the effects of helium (He) bubbles on the irradiation hardening. It was found that the degree of irradiation hardening of the two materials increased with an increase of ions fluences. The hardening range of SLM 316L SS was 6-47% while that of TM 316L SS was 32-75%. Moreover, the hardening saturation phenomenon was more pronounced in TM 316L SS than in SLM 316L SS. These results suggested that SLM 316L SS owned higher resistance to irradiation hardening than TM 316L SS. TEM observations showed that the density of He bubbles in TM 316L SS was 7.06 x10(23)/m(3) while that in SLM 316L SS was 3.33 x10(23)/m(3). Further, the average sizes of bubbles were 2.2 and 2.4 nm, respectively. The differences of the number density might be why SLM 316L SS had a slighter irradiation hardening than TM 316L SS, which was consistent with Orowan model. It was suggested that the numerous intrinsic microstructures such as entangled dislocation network walls and subgrain boundaries in SLM 316L SS could greatly reduce the density of He bubbles. (c) 2021 Elsevier B.V. All rights reserved.

In reply to the Comment by O. Neufeld et al. [Phys. Rev. A 105, 047101 (2022)], we argue that the conclusions of Phys. Rev. A 103, 043106 (2021) remain valid. We disprove the claim that the unphysical even-order harmonics originate from convergence issues related to reflections at the boundary of the simulation box. By additional calculations, we show that such reflections perturb the high-order harmonic generation spectra by oscillations with periods much smaller than the distance between the harmonics. We also demonstrate that the convergence argument of the Comment, in contrast to our multielectron excitations argument, cannot explain why there are no unphysical even-order harmonics in one-electron systems. Moreover, we show that the argument put forward in the Comment to conclude that the time-dependent Kohn-Sham equations are superior to the time-dependent natural Kohn-Sham equations is not valid.

This study aims to provide microstructural characterization for the matrix graphite which molten salt reactors (MSRs) use, and improve resistance to molten salt infiltration of the matrix graphite for fuel elements. Mesocarbon microbeads (MCMB) densified matrix graphite A3-3 (MDG) was prepared by a quasi-isostatic pressure process. After densification by MCMBs with average particle sizes of 2, 10, and 16 μm, the pore diameter of A3-3 decreased from 924 nm to 484 nm, 532 nm, and 778 nm, respectively. Through scanning electron microscopy, the cross-section energy spectrum and time-of-flight secondary ion mass spectrometry, resistance levels of the matrix graphite to molten salt infiltration were analyzed. The results demonstrate that adding a certain proportion of MCMB powders can improve the anti-infiltration ability of A3-3. Meanwhile, the closer the particle size of MCMB is to the pore diameter of A3-3, the smaller the average pore diameter of MDG and the greater the densification. As a matrix graphite of fuel elements in MSR was involved, the thermal and mechanical properties of matrix graphite MDG were also studied. When densified by the MCMB matrix graphite, MDGs can meet the molten salt anti-infiltration requirements for MSR operation.

The Hastelloy N alloy, Alloy 800H and 316H stainless steel were irradiated by Xe20+ ion irradiation with energy of 4 MeV at room temperature (peak damage ranging from 0.5 to 10 dpa). The micromechanical properties, hardness and creep plasticity, of these three investigated alloys were characterized before and after irradiation using nanoindentation. The results show that the hardness increases, and creep plasticity degrades with increasing ion dose in all the samples. In comparison, Hastelloy N has good irradiation damage resistance, while that of the 800H and 316H alloys is slightly worse. Additionally, the approximate positive relationship between irradiation hardening and creep plasticity degradation means that the property of creep plasticity of irradiated materials can be reflected from the nanohardness measurement for the heavy ion irradiation cases.