[en] Highlights: • This paper reports the full characterization of dislocation loops in single crystal ThO₂ after 0.47 dpa proton irradiation at 600^oC. • Dislocation loops were found to be mainly edge type with different variants of faulted $1/3〈111〉$ Frank loops along $\left\{111\right\}$ habit plane. • Inside-outside contrast technique and atomic resolution STEM-HAADF imaging revealed the interstitial nature of dislocation loops. This work focuses on the full characterization of dislocation loops induced by proton irradiation in single-crystal ThO₂. Irradiation was performed using 2 MeV H⁺ ions with sample temperature at 600^oC and a dose of up to 0.47 displacements per atom (dpa). Transmission electron microscopy (TEM) characterization was performed on a large number of dislocation loops. Burgers vector $\left(\overrightarrow{b}\right)$ analysis using standard $\overrightarrow{g}.\overrightarrow{b}=0$ invisibility criterion revealed different variants of $1/3〈111〉$ type dislocation loops. TEM analysis of edge-on dislocation loops was used to determine habit planes as $\left\{111\right\}$ type. The nature of dislocation loops was revealed using the inside-outside contrast method as interstitial type. Rel-rod dark field images were obtained by selecting streak at $\overrightarrow{g}=1/2\left[31\overline{1}\right]$ in diffraction pattern to identify the Frank loops present in the microstructure. Subsequent analysis of a single dislocation loop using atomic resolution scanning transmission electron microscopy (STEM) confirmed the interstitial nature of the loop with $\left\{111\right\}$ habit plane.

[en] The physics-based fission gas behaviour model available in the BISON fuel performance code provides satisfactory predictive capabilities for application to light-water reactor conditions. In this work, we present a model extension for application to fast reactor (U,Pu)O${}_2$ fuel. In particular, we detail the introduction of a lower bound to the number density of grain-face bubbles, representing a limit to the coalescence process once extensive bubble interconnection is achieved. This new feature is tested first against an experimental database for UO${}_2$-LWR, and secondly is validated against integral irradiation experiments for fast reactor (U,Pu)O${}_2$ fuel rods irradiated in the FFTF (Fast Flux Test Facility) and in the JOYO reactors. The comparisons of BISON results with the experimental data are satisfactory and demonstrate an improvement compared to the standard version of the code.

[en] High thermal conductivity additives are being considered to create composite nuclear fuels with a higher effective thermal conductivity (ETC) to reduce the peak fuel temperatures during reactor operation. However, the benefits of the additive may be reduced, or possibly eliminated, when placed into a reactor environment. In this work, we investigate the impact of fission gas bubbles on the ETC of composite fuels. We create 2D simulations representing a small portion of a composite UO${}_2$ fuel containing a single additive particle and calculate the ETC with various volume fractions of fission gas bubbles. Our results show that fission gas bubbles, depending on the volume fraction and contact angle at the additive interface, can completely remove any benefit of the additive and even result in a lower thermal conductivity than in UO${}_2$ without the additive. We also expose a trend that relates the fraction of additive that is screened by bubbles to the ETC. We then propose a model which predicts the ETC of a composite fuel with fission gas bubbles to better estimate the benefits of a novel fuel design.

[en] As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. Here, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr₂ phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)₆Fe phase (space group I4/mcm) and cubic (U,Zr)(Fe,Cr)₂ phase (space group Fd$\overline{3}$m). The (U,Zr)(Fe,Cr)₂phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm$\overline{3}$m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4₂/mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

[en] Morphology is a fundamental attribute when investigating voids and bubbles in ${\text{UO}}_2$. This study uses molecular dynamics and Monte Carlo simulations to predict the lowest energy shapes for voids and bubbles in ${\text{UO}}_2$. The energies of the $\left\{100\right\},$ $\left\{110\right\}$ and $\left\{111\right\}$ surfaces have been calculated and used to predict the equilibrium void shape from Wulff construction. This equilibrium shape is compared to low energy faceted voids exhibiting different relative proportions of each family of terminating surfaces. It is found that the equilibrium Wulff shape does not represent the lowest energy morphology for nm void sizes at temperatures between 300 K and 1200 K. Furthermore, the lowest energy faceted voids are slightly more energetically favourable than spherical voids, and as Xe is added, and bubble pressure increases, the faceted morphology becomes even more favourable than the spherical shape.

[en] Density functional theory (DFT) calculations of the thermodynamic and kinetic properties of point defects in the $\beta$ phase of uranium are reported. Defect energies and entropies were calculated using $2\times 2\times 2$ supercells and the Generalized Gradient Approximation (GGA) of Perdew-Burke-Ernzerhof (PBE) for the exchange-correlation potential. Due to computational cost, calculations of the vibrational properties governing entropies were performed by only displacing atoms within (roughly) the 3rd or 4th nearest neighbor shell of the defect, which implicitly assumes that atoms beyond this distance are unaffected by the defect. Migration barriers were estimated by nudged elastic band (NEB) calculations. The low symmetry of the $\beta$-U phase (the unit cell is tetragonal and contains 30 atoms) results in many point defect configurations and even more migration pathways. A connectivity map, starting from the most stable point defects, was developed in order to identify the rate-limiting step controlling the net diffusion rate in each crystallographic direction. The uranium self-diffusivity tensor was calculated by combining the defect formation energies, entropies, migration barriers and attempt frequencies. The fastest diffusion rate was determined to be a vacancy mechanism in the z crystallographic direction. The predicted uranium self-diffusivity for this mechanism agrees well with available experimental data. The diffusion mechanisms and rates identified in this study will inform fuel performance models of swelling and gas evolution.

[en] Materials properties of uranium oxycarbide (UCO), a fuel candidate for high temperature gas reactors (HTGRs), are investigated using the Hubbard corrected density functional theory (DFT+$U$) and ab initio molecular dynamics (AIMD+$U$). To accurately describe UCO systems within the supercell approach, we have evaluated the structural disorder of the ion sublattice within the special quasirandom structures (SQS) approach. The ground-state of UC_xO_2-x and UC_1-xO_x systems was found by examination of the electronic configuration, the occupation matrix and the Hubbard $U$ parameter. We obtained the electronic properties, lattice structure, formation enthalpy and free energy of mixing of the stoichiometric UC_xO_2-x and UC_1-xO_x phases. Subsequently, we have determined the preferential phase of UCO (UC_xO_2-x or UC_1-xO_x) for low carbon and oxygen concentrations, respectively. In particular, UC_xO_2-x was found stable for carbon content lower than 50%, with carbon atoms in the first coordination shell out of the perfect fluorite positions, driven by the formation of C₂ groups as in $\alpha$-UC₂. Moreover, the analysis of the radial distribution functions shows that structures with C₂ groups remain stable in the range of temperatures from 300 to $\sim$2500 K.

[en] Highlights: • The monotectoid transformation temperature of Zr-1Nb is increased by replacing Nb by Ta. • The enthalpy of formation of the ${\beta }_{Nb/Ta}$ phase increases with the added Ta. • A slight decrease in hydrogen solubility was observed at low Ta content. • The resistance to corrosion of Zr-1Nb is increased by replacing Nb by Ta. The feasibility of using Ta as a minor alloying element in Zr-1Nb alloys is addressed in terms of phase stability and resistance to corrosion of the microstructure obtained with the following two thermomechanical processes after a water quenching from ${\beta }_{Zr}$: a) annealed at 570°C for 3840 h and b) a combination of intermediate annealing temperatures and cold rolled steps. Hydrogen solubility was also measured, but only in alloys with the first process. It was found, through calorimetric measurements, that an increase of Ta replacing Nb in the nominal alloy composition increases the $\alpha \to \left(\alpha +\beta \right)$ phase transformation temperature in agreement with an estimated increase in the enthalpy of formation of the ${\beta }_{Nb/Ta}$ phase, $\Delta H_{{\beta }_{Nb/Ta}}^f$. As expected, the mass fraction of ${\beta }_{Nb/Ta}$ measured in the rolled alloys is lower. Further, this diminution is larger in the alloys containing Ta due to the lower mobility of Ta and higher $\Delta H_{{\beta }_{Nb/Ta}}^f$. The growth of the oxide layer in each alloy, at 400°C steam up to 14 days, follows the predictor calculated as the ratio between the atomic concentrations of Nb plus Ta in the alloy and $\Delta H_{{\beta }_{Nb/Ta}}^f$. A slight decrease in hydrogen solubility was observed at low Ta content in the alloy.

[en] The FeCrAl-oxide dispersion strengthened (ODS) alloy is a promising candidate alloy for the accident tolerant fuel (ATF) cladding of light water reactors (LWRs) and has been recently developed in Japan. This paper reports on the progress of the development of accident tolerant FeCrAl-ODS fuel claddings for boiling water reactors (BWRs) in Japan.

spara021^>Both experimental and analytical studies were conducted to evaluate the influence of implementation of the FeCrAl-ODS fuel cladding to the current BWRs.

spara022^>In the experimental study, key material properties of FeCrAl-ODS fuel cladding have been obtained and accumulated using bar, sheet and tube-shaped materials to support the evaluations in the analytical study. In the evaluation at normal operating conditions, strength and ductility, corrosion property, tritium permeability, wear property, resistance to iodine stress corrosion cracking (SCC) and weldability were examined. A preliminary assessment of compatibility of the FeCrAl-ODS fuel cladding with the current recycling system in Japan was also conducted. In the evaluation of the design basis accident and the beyond design basis accident, strength, steam oxidation property, resistance to water quenching during a loss-of-coolant accident (LOCA) and the LOCA burst property were examined.

spara023^>In order to evaluate the influence of implementation of the FeCrAl-ODS cladding to the current BWRs, the core characteristics and the fuel behavior were evaluated in the analysis study at the normal operating condition. The analysis for the 9 × 9-type and the 10 × 10-type fuel assemblies and the reactor type of the Advanced BWR (ABWR) revealed a good applicability of FeCrAl-ODS fuel cladding.

spara024^>Finally, the challenges and perspectives found in the program are highlighted to enhance international collaborations to promote the development of the FeCrAl-ODS fuel cladding.

[en] Highlights: • Incorporation in a glass of a dismantling waste rich in Zr, P and Mo • Impact of waste composition variation and waste loading on glass microstructure • P and Mo are the only elements that may lead to phase separation or crystallization • For Mo-rich compositions Cs may enter into molybdate crystals (CsLiMoO₄) Dismantling nuclear facilities leads to radioactive waste less active but which may have highly variable compositions compared to the high-level radioactive wastes recovered after the reprocessing of spent nuclear fuel. In this work, we studied the ability of an alkali-rich glass matrix belonging to the SiO₂-B₂O₃-Al₂O₃-Fe₂O₃-Na₂O-Li₂O-CaO system to solubilize P₂O₅, MoO₃, ZrO₂ and Cs₂O by melting at 1100°C. Phosphorus, molybdenum, zirconium and cesium are present as a mixture of complex compounds in the real radioactive dismantling waste containing ¹³⁷Cs considered here. To determine the capacity of the matrix to accept a wide range of variations of waste composition and the solubility limits of P₂O₅, MoO₃, and ZrO₂, several glass series were prepared by melting mixtures of raw materials (oxides, carbonates, phosphates) and by increasing the total amount of oxides representing the waste and varying their relative proportions. Their incorporation in the melt was studied by analyzing the microstructure of quenched glasses by XRD and SEM-EDS. In addition, the phase separation and crystallization tendencies during cooling were studied by analyzing the microstructure of glasses cooled at 1°C.min^-1 from 1100°C (representative of cooling in industrials steel canisters). It is shown that the glass can accept a wide range of waste compositions without exhibiting heterogeneities. For all compositions the melt remained homogeneous (study of quenched samples). However, during slow cooling, P₂O₅ and MoO₃ may lead to phase separation and crystallization of Na₂MoO₄, CsLiMoO₄, NaCaPO₄, NaLi₂PO₄, and Li₃PO₄. Cs can be partially incorporated into the molybdenum-rich phase CsLiMoO₄ when MoO₃ content is higher than 1.3wt%. ZrO₂ never lead to phase separation or crystallization, possibly because of the existence of strong connections between Zr and Si through Zr-O-Si bonds whereas P and Mo would be present as PO₄³⁻ and MoO₄²⁻ mobile entities. The increasing order of oxides solubility in the glass is the following: MoO₃2O₅2.