The effect of Au-ion irradiation on the morphology, microstructure and lead-bismuth eutectic (LBE) corrosion behavior of refractory TiNbZrMoV high-entropy alloy coating was investigated. After irradiation, the surfaces of coatings were smoothened, and crystalline phase was formed in the coatings. After LBE corrosion at 650 °C, the irradiated samples presented a triple layer (interface Fe-Cr spinel, coating and surface Fe-oxide layer) consistent with the pristine sample, and the coating still maintained the low-consumption characteristic. Irradiation accelerated LBE corrosion, which could be attributed to the formation of grain boundary and defects that provided more diffusion channels for atoms.

Abstract not available

The oxidation behaviour of uranium nitride (UN) powder has been examined while allowed to age over a period of 129 days with both oxidation in synthesised air and high temperature water observed. The UN powder was stored in air and in an argon filled glove box (<3ppm O2) to determine if the storage environment would affect the oxidation. Within this dataset, no significant change to the oxidation onset temperature or reaction kinetics was observed outside of measurement uncertainty with respect to the age of the powder.

F82H steel and pure Cr metal were joined by solid-state diffusion bonding under high vacuum at various temperatures below Ac1 in the range 964 to 1072 K for 240 min. Bonding was successful at all temperatures and the diffusion behaviour, chemical composition and microstructure at the F82H/Cr interface and interdiffusion zone were characterised using SEM-EDS and TEM analysis. A hard, M23C6-rich interface layer formed on the Cr side of the joint due to diffusion of C from the F82H steel. The thickness of the M23C6-rich interface layer reduced with reducing temperature, from ∼1.1 µm when bonded at 1072 K, to ∼0.2 µm at 964 K. Cr diffusion into F82H was only observed at the highest temperature of 1072 K, with minor Cr enrichment detected at a depth of <1 µm into the F82H. The slow diffusion behaviour below Ac1 meant a lath martensite microstructure appears to have been retained throughout the F82H after bonding below the typical tempering temperature, and hardness testing of the F82H suggests no post-bonding heat treatment would be required. Simple models for the interface evolution of the F82H/Cr joint both above and below Ac1 are proposed, which can act as reference points for the further development of Cr-coated F82H and its likely behaviour in a range of fusion power plant operational scenarios.

Uranium borides are currently being investigated as an accident tolerance fuel or advanced technology fuels (ATFs) to replace conventional UO2 fuel due to their high uranium density and high thermal conductivity. The surface properties of diverse facets of a crystal are pivotal for understanding different corrosion behaviors, fission gas bubble behaviors, and equilibrium morphologies. In this study, we employ density functional theory (DFT) calculations to investigate the surface properties of uranium borides. The surface orientations with maximum miller index up to 4, 1, and 2 for UB2, UB4 and UB12 were analyzed, respectively. Based on the calculated surface energies, the other surface properties of uranium borides were obtained through Wulff construction, including equilibrium morphology, dominant surface orientations, area-weighted surface energy and work function. Additionally, grain boundaries (GBs) have a significant impact on material properties, such as mechanical properties, corrosion behavior, and resistance to irradiation. Therefore, we use DFT calculations to determine the GB energies and work of separation for three low-Σ (7) symmetrical tilt GBs (STGB) and two twist GBs (TGB).

Studies on high burnup UO2 subjected to loss-of-coolant accident conditions have shown that restructured regions of the fuel are susceptible to pulverization and eventual dispersal. Due to a lack of pre-test characterization, the distinct microstructural features rendering the fuel prone to fragmentation remain ambiguous. Four samples of commercially irradiated light-water reactor UO2 have been characterized utilizing electron backscatter diffraction to assess the susceptible microstructure. The microscopy focused on determining the burnup and temperature conditions responsible for the formation of the different microstructural regions where the regions were denoted as the high-burnup structure (HBS), HBS transition, mid-radial, restructured central, and central region. Previous works have outlined the specific conditions required for the restructuring of the microstructure into the HBS, but the conditions responsible for the restructuring in the central region of the fuel are not well understood. The four analyzed samples confirm a burnup threshold of 61 GWd/tU, and an unknown temperature range is needed to facilitate the formation of the restructured central region. Additional fuel performance evaluations are needed to quantify the temperature range promoting restructuring in the central region.

Li-D films co-deposited from deuterium plasma were investigated. It was found that the D concentration has the dependence on the deposition temperature with a maximum of ∼30 at. % at 350–500 K. Lithium mass deposited on the substrate decreases rapidly at deposition temperatures above 570 K. In the narrow range of 520–570 K lithium collection is still efficient and deuterium accumulation is already low, so the temperatures in the range are optimal for lithium collectors in fusion devices. Deuterium in the films is mainly in the form of lithium deuteride with small amounts of D is solution and in lithium hydroxide and lithium hydronitrides.

Carbon steels have been extensively used as structural materials in the secondary side of pressurized water reactors, which are susceptible to flow-accelerated corrosion (FAC). To overcome this issue, low-alloy steels containing 2.25Cr-1Mo (P22) have been introduced, as Cr and Mo are effective alloying elements in mitigating FAC. However, the behavior of these alloying elements in the secondary water chemistry is not well-understood, and therefore, there is a need to explore alternative materials that can address the issue of FAC. In this study, three model alloys were manufactured based on Ducreux's model on FAC rate to examine the effects of Cr and Mo on the corrosion behavior of low-alloy steels compared to that of commercial P22. Microstructure of P22 was ferritic/pearlite structure while the model alloys was ferritic/tempered bainite structure due to the existence of Mn and Si. The difference in microstructure led to the difference in hardness values, and elimination of Mo from the model alloys resulted in reduction in strain. The FAC rate was mainly influenced by the Cr content in the oxide layer rather than Mo. The passive film layer hinders the exposure of the inner layer to H2O in the solution, resulting in Cr3+ dissolution into the Fe oxide layer. The continuous passivation leads to the formation of two compact layers, one amorphous and one crystalline Cr, creating Fe oxide substitutes. It is also revealed that the solubility of Cr species is much lower than that of Fe species resulting in the enrichment of Cr in the outer layer, and thus the higher Cr content reinforces the passivity of the steels. The Fe-Cr alloy exhibited promising corrosion resistance, suggesting that it could be a potential substitute for Fe-Cr-Mo alloy.

Fully Ceramic Microencapsulated (FCM) fuel is a new type of material in nuclear energy field. It has become a promising accident tolerant fuel (ATF) candidate due to its advantages of high radiation stability, high fission product containment capacity and excellent thermo-mechanical performance. During operation, the temperature, stress, deformation and other key parameters will change significantly, which directly affect the safety of nuclear reactor. Therefore, it is necessary to carry out a multi-physics fully-coupled analysis suitable for the multi-scale, multi-component, multi-phase and nonlinear coupling processes in the reactor, so as to sort out the influences of material properties, radiation behavior, thermo-mechanical performance from numerous influencing factors. In this paper, based on the irradiation-thermo-mechanical coupling method, the performance of TRISO-based FCM fuel under irradiation conditions was analyzed, key thermo-mechanical parameters and other influencing factors of the fuel were obtained, failure behavior of the fuel element was evaluated. On this basis, the theoretical and numerical model of multi-objective optimization method were established, the optimum parameters including material properties, fuel geometry and particle arrangement were found by optimization method. The optimization results showed that the multi-objective optimization results could provide multiple Pareto optimal solutions for the FCM fuel. Specifically, when the TRISO particle spacing is 200 µm and 150 µm, the objective function difference between the optimal heat transfer solutions is only about 1.7%, meanwhile the mechanical properties vary significantly (the difference between the objective functions of the optimal solutions is about 12.4% ∼26.5%), indicating the response of stress is more sensitive than temperature. When the TRISO particles are tightly packed (particle spacing less than 100 µm), the difference of objective function between the optimal solutions is less than 5%, illustrating the optimization has limited improvement on the thermo-mechanical performance of the fuel, namely the optimization space is limited.

Borosilicate waste glasses were successfully developed to immobilize two high-level waste raffinate streams via an in-can melter process with an Inconel 601 canister at 1050 °C. Measured viscosity and crystallinity thermal profiles were within the targeted processing constraints for the in-can melter process. Measured chemical durability of the glass by ASTM C1285–21 (Method A), ranged from normalized loss of boron, NL(B) = 1.44 – 2.65 g·m−2, and NL(B) decreased with increased waste loading, accompanied by increased SnO2 crystallinity. Measured corrosion of the in-can melter canister by a glass melt showed that Inconel 601 performed well at 1100 °C for up to 500 hr. Resistance polarization measurements versus time revealed that Inconel 601 corrosion rates in (and by) glass melts decreased from an initial rate of 63 mm·y−1 down to 10.2 mm·y−1 after 137 h with increased duration, which was attributed to formation of an oxide passivation layer (mainly Cr2O3) at the alloy-glass interface.

Tungsten (W)-based high-entropy alloys (HEAs) have shown promising properties as nuclear fusion materials. Characterizing the primary damage is a critical step in describing and revealing the irradiation-induced damage and radiation resistance mechanism. However, there are a limited number of studies on collision cascades and primary radiation damage in W-based HEAs, owing to the large amount of calculations involved and a lack of appropriate interatomic interaction potentials. In this work, we developed a semi-empirical interatomic potential for W–Ta–Cr–V. By using the developed potential, molecular dynamics simulations of collision cascades in W-based HEAs were performed to assess the primary damage due to irradiation. Based on experimental samples, we reported defect production in W38Ta36Cr15V11 and compared it to pure W for primary knock-on atom with energies ranging from 1 keV to 100 keV. Our findings showed that the number of FPs at the thermal spike and the number of surviving FPs at the end of the cascade in W38Ta36Cr15V11 are more than those in pure W, mainly due to the lower threshold displacement energy, melting temperature, and formation energy of point defects. Collision cascades in the W38Ta36Cr15V11 are less likely to result in the formation of dislocation loops compared to pure W. After collision cascade, in W38Ta36Cr15V11, the concentrations of Cr and V atoms in defects is significantly higher than their corresponding concentrations in the system, showing an aggregation tendency. The current collision cascade results provide insights into the primary damage of W-based HEA system under irradiation and should provide reliable guidance for describing the primary damage source terms needed in the kinetic models used to simulate radiation-induced microstructural evolution.

In this paper, we show how results of multiscale modeling can be condensed to establish a simple analytical flow equation for irradiated low-alloy steels. The flow equation accounts for temperature, strain rate, the initial and irradiation microstructure. Starting from a complete set of constitutive equations describing crystal plasticity of irradiated ferritique steels, simple assumptions and approximations are used to integrate the equations over the homogenized polycrystal. The final flow equation incorporates the lattice friction resistance, the Hall-Petch effect, the forest strengthening and irradiation hardening. Every component is established by multiscale modeling at lower scales. Irradiation hardening is attributed to solute clusters. Their size and density are determined in experiment, their shear resistance computed by atomistic simulations and their contribution to the flow stress assessed using dislocation dynamics simulations. The formation of solute clusters is found to increase the flow stress and the strain hardening modulus. Irradiation hardening decreases with temperature and with the yield stress of the unirradiated material. Predictions of the tensile curves and irradiation hardening are compared with many experimental results. A special section is dedicated to the interpretation of irradiation hardening measured in a surveillance program.

This study deals with the assessment of tensile properties across the nozzle to primary piping safe-end employing the instrumented indentation testing (IIT). Standard tensile tests were performed in order to validate the results of IIT. Tensile properties were also estimated from conventional hardness test. Both the yield strength and tensile strength measured by IIT were in satisfactory agreement with the results of standard tensile tests and were found superior compared to the values estimated from the conventional hardness measurement. Highest deviation between the results of IIT and tensile tests was found in the regions of weldments. IIT showed as capable of measuring the variation of tensile properties across the pressure vessel nozzle to primary piping safe-end. Moreover, IIT has the advantage over the tensile testing that specimens do not need to be extracted from the studied piece and more detailed distribution of mechanical properties can be acquired.

In materials exposed to neutron irradiation, helium atoms produced by transmutation reactions are prevalent and play a significant role in the evolution of dislocation loops, causing mechanical property degradation. However, systematic investigation on the helium role is lack and possible origin of helium effects is unclear. Here, by performing 400 keV Fe ion and 30 keV He in-situ irradiation, this study found that about 9.2×1012 cm−3 loops nucleated for every increase of 1 appm He at 0.9 dpa in Fe++He+ dual beam irradiation and 87% loop density was produced with the help of helium at 2262 appm/dpa. The different loop evolution in single He irradiation and Fe++He+ irradiation indicated the presence of high-density TEM-invisible (≤ 2 nm) small loops after Fe+ irradiation at high temperatures. The large number of loops produced by helium irradiation were possibly related with the small bubbles.

The pre-disposal management of Radioactive Liquid Organic Waste (RLOW) is hampered by its challenging physico-chemical properties. In this work, a straightforward conditioning option based on RLOW impregnation on absorbing materials and followed by encapsulation in a stable geopolymeric matrix is proposed, avoiding onerous pre-treatments and the use of surfactants. Recycled materials have been investigated as adsorbent and geopolymer precursors to foster process sustainability. Relevant properties have been studied to ascertain the waste acceptance criteria accomplishment: materials compatibility, RLOW loading factor and bleeding, microstructure, compressive strength, leaching and thermal stability. This approach is promising, although some criticalities remain unsolved.

Molten salts have many applications ranging from a heat transfer medium in both generation IV nuclear reactor designs and the solar industry to thermal storage systems. While molten salts show promising properties for these applications, there still exists a knowledge gap for the transport properties of molten salts at elevated temperatures. This work uses ab initio Molecular Dynamics to investigate the transport properties of KCl, LiCl, KCl-LiCl eutectic, NaCl, MgCl2, and NaCl-MgCl2 eutectic molten salt systems. The properties presented here are the diffusion coefficient, viscosity, and isochoric heat capacity. These properties are compared to experimental data where available and other computational work in cases where no experimental data is available. This is the first work to explore timescales over 100 ps via AIMD for the determination of transport properties in molten salts.

The thermal decomposition of Pu(III) oxalate was analyzed by Raman microspectroscopy, infrared spectroscopy, scanning electron microscopy, and powder X-ray diffraction. These data show that crystalline Pu2(C2O4)3•9H2O progressively loses water and oxalate ligands as it is heated, which leads to a decrease in long-range lattice ordering, though minimal changes are observed in gross crystalline morphology. The onset of PuO2 formation was observed between 200 - 250 ℃. Thermal decomposition of oxalate ligands leads to the formation of CO2 and plutonium oxalate-carbonate moieties, which had not been observed in previously published thermogravimetric measurements of Pu(III) oxalate. Formation of plutonium oxalate-carbonate moieties is believed to be associated with a change in the plutonium oxidation state from 3+ to 4+, which occurs prior to PuO2 formation. The data provided herein demonstrate the rich spectroscopic nature of a rather underexplored, and technologically relevant, plutonium system. Ideally these results will further future investigations into the Pu(III) oxalate system both experimentally and computationally.

The influence of ion-incident free surfaces on void swelling in near-surface regions was systematically studied using self-ion irradiation of single crystal pure iron. The key interest was to evaluate the effect of free surfaces not only on the formation of void-denuded zones but also on the swelling in the region immediately adjacent to the void-denuded zone. The irradiation matrix included beam energies varying from 1.0 to 5.0 MeV, peak displacements-per-atom levels of 50 and 100 dpa, and irradiation temperatures at 425 °C (at 6 × 10−3 peak dpa/sec, 475 °C (at 1.2 × 10−3 peak dpa/sec) and 525 °C (at 6 × 10−3 peak dpa/sec). The observed denuded depth Δx obtained from transmission electron microscopy was modified to incorporate the sputtering effect. The derived activation energy governing the denuding process is Ex4=1.65±0.03eV, which is significantly higher than the vacancy migration energy EVm known for Fe to be 0.67 eV. This difference is speculated to be due to the strong effect of relatively small amounts of dissolved carbon at 103 appm, which reduces the effective vacancy mobility and thereby increases the effective migration energy. Based on the effective vacancy migration energy extrapolated from a previous first-principle calculation for Fe containing various carbon levels, rate theory calculation was used to model the surface influence and the results support the speculation of carbon influence. For the region immediately adjacent to the void-denuded zone, swelling is locally suppressed, rising to reach the bulk swelling at a depth increment essentially equal to the denuded width. This suppression effect causes shifting of swelling vs. local dpa curves obtained from different peak dpa irradiations. Such shifting is proposed in the present study as a method to define the boundary of the surface-induced suppressed swelling zone beyond which the swelling can be considered to be fully free of surface influence.

Lithium-based ceramics exhibit good thermophysical properties, have low activation and chemical activity, and also release tritium well. These characteristics make lithium ceramics the best candidates for use as a functional material for a solid breeder blanket. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials can only be obtained from the results of reactor experiments. On the other hand, many important parameters can only be estimated in post-irradiation experiments (PIE). This paper describes studies of tritium and helium release in post-irradiation experiments (PIE) on the thermal desorption from ceramic pebbles Li2TiO3 + 5mol% TiO2 (with 96% enrichment on lithium-6 isotope), irradiated in the WWR-K reactor for 223 days. In PIE a peak of helium release was recorded for each sample in the temperature range of 1300–1500 K. An assumption was made in the paper to explain the nature of this helium release peaks from lithium ceramic pebbles.

The Brailsford-Bullough model of recombination of point defects on the surface of coherent inclusions under irradiation is critically analysed. It is shown that the steady state solutions of the model exist only for relatively small coherent particles (of a few nanometres in size) and, depending on the configuration of the particle-matrix interface, at relatively low temperatures. For high temperatures, the model should be reformulated, taking into account thermal desorption of trapped vacancies, which makes it possible to explain the stabilization of trapped point defects in the steady state and to calculate the rates of removal of point defects by coherent inclusions.