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Uranium plays an unquestionable role in the framework of nuclear physics, biology, and radiopharmacy. Moreover, uranyl ion UO22+ offers an immense variety of applications due to the unique photosensitivity of its complexes. The excited state of uranyl cation is indeed accessible under ultraviolet‐visible (UV–vis) light, readily producing radical species UO22+* upon light irradiation. Herein, an innovative synthesis protocol is presented to explore the use of uranyl cations as photocatalyst systems for photocurable sol–gel‐based formulations, coupling the photochemical reactions of uranyl cations with photopolymerization‐based additive manufacturing processes. Additive manufacturing has nowadays revolutionized the production of complex structures with arbitrary geometries and has opened up enticing opportunities for innovative technological breakthroughs and highly tailorable systems. The fabrication of micro‐architected components is shown via vat photopolymerization, namely, the Digital Light Processing technique, and ‐3D) printed parts are converted into uranium dicarbide (UC2)/carbon nanocomposite upon carbothermal reduction. This uranyl‐mediated additive manufacturing process constitutes the first application of the synergistic role of uranyl motifs in a photopolymer platform, demonstrating for the first time the possibility to directly pattern uranium‐based materials in complex structures. [ABSTRACT FROM AUTHOR]

Canada has operated 17 research reactors at 11 different locations. The spent fuel from these research reactors differs significantly from CANDU fuel, which makes up the vast majority of spent fuel in Canada, and will eventually require disposal. The focus of this paper is to identify properties specific to Canadian research reactor fuel designs that would impact their suitability for disposal. The radionuclide inventory, hazardous chemical inventory, decay heat, residual enrichment, radiation decay rates, and gas generation of several Canadian research reactor fuel designs were simulated using the SCALE 6.2.4 software suite. The National Research Universal U3Si/Al dispersion rod, the National Research Experimental uranium metal X-rod, the Royal Military College UO2 SLOWPOKE-2 core, and the Whiteshell Reactor 1 uranium carbide bundle were investigated. Fuel burnup is the primary driver for the concentration of most radionuclides, hazardous chemicals, decay heat, and radiation decay rates. Carbon-14, chlorine-36, and mercury are driven by initial impurities in the fuel, which vary by fuel design. Low burnup, enriched fuels constitute a reasonable bounding case for the evaluation of criticality safety and proliferation risks. Canadian research reactor fuels are unlikely to present a greater risk of over pressurization from helium generation than CANDU fuel. Overall, the small volume of Canadian research reactor fuels requiring disposal is an important factor in the evaluation of disposal requirements. [ABSTRACT FROM AUTHOR]

In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2. [ABSTRACT FROM AUTHOR]

The use of uranium carbide for extracting uranium, a rare and valuable material, has garnered significant research interest. The long-term storage of uranium scrap is challenging due to the risk of ignition and contents of nuclear materials. Thus, scrap treatment is required for safe storage and volume reduction. We investigated the oxidation behavior of different types of multi-walled carbon nanotube-containing uranium scraps (green and sintered) in different forms (fragmented and powdered) using thermogravimetric/differential thermal analysis and volatilization studies. An air flow and oxidation temperature of 500 cm3/min and 900 °C, respectively, were suitable for the effective recovery of inert U3O8 for safe storage or recycling. For optimal results, scrap in powder form should be used. These findings provide insights for recycling uranium from uranium carbide scrap, with potential applications in research and industry. [ABSTRACT FROM AUTHOR]

• Homogeneous model for FCM fuel analysis was examined. • We evaluated the performance of the SMART reactor using FCM and traditional UO 2 fuel loadings. • The study improved thermal conductivity and safety performance by using UC FCM and various cladding material combinations. • The optimum cladding design in comparison to reference cases and other variations of FCM-UC were examined. • We used MCNPX 2.6 code to perform the neutronic assessment of the SMART core, which revealed the appeal of FCM-UC fuel. Fully Ceramic Microencapsulated (FCM) fuel is one of the most advanced fuel types for the next nuclear reactor generations. The use of uranium carbide fuel in FCM offers multiple benefits such as high fuel density, improved thermal conductivity and enhanced mechanical stability. Both UC and conventional fuel (UO 2) have the same melting point. As a result, under reactor operating conditions, UC has better safety margins. In addition, FeCrAl is a promising cladding material. However, because of its greater thermal neutron absorption cross section, FeCrAl has a neutronic penalty when compared to Zircaloy leading to an annual reload requirement with high fuel enrichment. Owing to their low power density and excellent design, small modular reactors (SMRs) have exceptional inherent safety. In this work, we determined the optimum design of an accident tolerant fuel (ATF), with respect to neutronics, in SMART reactor using dense UC FCM (in contrast to the conventional UO 2 fuel) and a combination of zircalloy and FeCrAl (FCM -Zircaloy, FCM -FeCrAl, FCM -combining of zircaloy and FeCrAl) as cladding to achieve better thermal conductivity as well as safer performance, while considering the required fuel enrichment for the proposed designs. The effective multiplication factor and burnup were calculated for neutronic assessment of the SMART core, the outcomes are showcased alongside a comprehensive assessment of the characteristics and prospective. The neutronic parameters were estimated using the stochastic MCNPX 2.6 code. We found that Fully Ceramic Microencapsulated -Uranium Carbide (FCM-UC) fuel type was more attractive than conventional UO 2. The optimum cladding design as compared to the reference case and the other cases of (FCM-UC), was determined. When 50% FeCrAl cladding was employed, case 15 in our study emerged as the superior design, demonstrating maximum production of 239Pu due to higher initial concentrations of 235U and quantities of 238U. This work may offer recommendations for future optimized designs of the SMART reactor parameters. [ABSTRACT FROM AUTHOR]

Fuel performance analysis was conducted for silicon carbide (SiC) composite clad uranium carbide (UC) fuel of a 500-MW(thermal) gas-cooled fast reactor, specifically the energy multiplier module (EM2) under normal operation. The analysis consists of two parts: Part I includes a description of design bases and criteria, fuel element design specifications, and material properties and models, while Part II (this paper) includes the fuel modeling approach, computer code, and the fuel design evaluation. In Part II, the FRAPCON-4.0 code was updated to include material properties and models of UC fuel, SiC composite cladding, and helium coolant, and named FRAPCON-4.0GA. The analysis was performed using the hot rod power envelope and burnup history. The results show that the present design of the EM2 fuel element has ample margin to melting owing to the high thermal conductivity of the UC fuel and annular pellet configuration. The operating temperature of the fuel element also minimizes the radiation-induced deformation of the SiC composite cladding. The simulation results show that the hoop stress of the cladding is below its tensile stress limit, i.e., one-third of ultimate tensile stress, while the cladding hoop strain limit is reached at 22.5 year, which is less than its design life of 32 years. However, sensitivity calculations of the swelling rate and design parameters indicate that it is feasible to reduce the cladding hoop strain by accommodating the fuel swelling into the open pore. Considering uncertainties associated with the material properties and models, it is highly recommended to experimentally verify the UC swelling and SiC composite creep, which are critical properties in analyzing the long-life fuel behavior. [ABSTRACT FROM AUTHOR]

Fuel performance analysis was conducted for the silicon carbide (SiC) composite clad uranium carbide (UC) fuel of a 500-MW(thermal) gas-cooled fast reactor, specifically the energy multiplier module (EM2) under normal operation. The analysis consists of two parts: Part I (this paper) includes a description of design bases and criteria, fuel element design specifications, and material properties and models, while Part II includes the fuel modeling approach, computer code, and fuel design evaluation. In Part I, the design bases and criteria describe the maximum allowed material temperature, cladding stress limit for structural integrity, and cladding strain limit for hermeticity. The material properties and models have been collected from open literature and recent measurements for the UC and SiC composites, respectively. As a result of reviewing legacy UC properties and models, it is recommended to measure the as-fabricated EM2 fuel properties with high priority to the thermal conductivity, swelling rate, and mechanical strength. For the SiC composite cladding, it is recommended to refine the creep rate for its temperature and time dependence. The stress-strain model also needs to be refined for its strain rate, irradiation, and temperature dependence. [ABSTRACT FROM AUTHOR]

Gaseous volume oxidation (voloxidation) of uranium mononitride (UN) and uranium monocarbide (UC) was investigated in the "NOx-air" atmosphere in the temperature range 298–673 K. It was shown that UN can be converted into uranium water-soluble compounds using an alternative oxidation atmosphere based on NOX-gases. UO3 and uranyl nitrates are the main products of the reaction. Maximum degree of the UN conversion to water-soluble compounds equal to 80 % was observed at the temperature 565 K. Products of the UC conversion are uranium oxides and oxihydroxides. The observed degree of the UC conversion into water-soluble compounds was less then 20 %. [ABSTRACT FROM AUTHOR]

To study the influence of the structural properties of UC x targets on their release properties, several types of targets using different precursors (carbon and uranium) were synthesized, characterized, irradiated and heated leading to the determination of the released fractions of eight elements. In this article, the production rates of these targets are estimated under the use conditions at ALTO, i.e. with targets bombarded by an electron beam (10 μA, 50 MeV). We have simulated the fission number produced using the FLUKA code. Then, we have determined the release efficiency as a function of the half-life of the isotopes using average diffusion coefficients deduced for the elements studied previously. Finally, we compare the production rates obtained from the various targets and conclude that the target must be adapted to the element studied. It is crucial to find in each case the best compromise between the target density and the release efficiency. [ABSTRACT FROM AUTHOR]

Oxidation of uranium carbide (UC) small fragments from sintered pellets was experimentally tested to better understand UC safe-handling procedures given the renewed interest of non-oxide fuels for high temperature gas or liquid metal cooled reactors. Transformation from UC to U 3 O 8 via a self-ignition reaction was observed at partial pressure of oxygen as low as 10 Pa. The heat output from UC self-ignition in fragments (not-free from UO 2 contamination) previously stored in either air atmosphere or inert-atmosphere during a three months period was monitored at 973 K and 1073 K in air atmosphere in a TGA/DTA and no difference could be observed. Residual carbon content, measured as amorphous carbon, carbide or CO/CO 2 , decreased with exposition temperature in U 3 O 8 resulting oxide, this is in contrast with zirconium carbide resulting oxide, ZrO 2. Cracking and stresses accumulated in the oxide were highest for UC to U 3 O 8 compared to UC to UO 2 reactions and ZrC to ZrO 2 reactions. [ABSTRACT FROM AUTHOR]

The steel specimens were tempered in contact with uranium carbide powder by sodium bonding at 500, 600, 700, and 800 °C for 1000h. Carburizing zone of the specimens was determined by measuring of microhardness which is taken as a kinetic variable instead of the corresponding carbon content. Arrhenius equation was determined for the diffusion of carbon atoms in the steel by using the solution of Fick's second law. Temperature dependency of the activation enthalpy, Gibbs energy, and entropy was calculated from the transition state theory by an assumption that the carburizing occurs over an activated complex. Kinetic and thermodynamic for formation an activated complex were discussed depending on the obtained numerical values.

This project investigated the use of uranium nitride (UN) and uranium carbide (UC) reactor fuel and compared their performance to uranium oxide (UO2) in a nuclear reactor for space-based applications. As a baseline for analysis, the Prometheus Project reference reactor module was considered: a gas-cooled fast reactor using highly enriched UO2 fuel with 1 MW of thermal power output and a 15-year core life. An estimate of the temperature feedback effect on reactivity was made for each fuel type at the beginning, middle, and end of core life; results for each fuel were compared. This analysis indicates that UN-fueled reactors may exhibit a stabilizing negative reactivity feedback for increasing temperatures and that this benefit persists in the face of fuel composition changes over core life. The benefit of increased uranium loading density was assessed through a quantitative estimate of overall core weight for each fuel. It was found that weight savings on the order of 1000 kg can be realized for a reactor of this size by using either UC or UN rather than UO2. [ABSTRACT FROM AUTHOR]

Highlights • Lowering temperature from 1173 K to 873 K improved conversion of UC pellets to U 3 O 8. • Slower oxidation at 1173 K were related to the formation of a thick oxide layer. • Temperature influenced UC oxidation in a similar way in experiments performed in 1 atm air in furnaces and in 10 Pa O 2 in a High Temperature Environmental Scanning Electron Microscope (HT-ESEM). Abstract Oxidation of UC was studied from 873 to 1173 K in air and in 10 Pa oxygen using a High Temperature Environmental SEM (HT-ESEM). Conversion to U 3 O 8 improved when using 873 K as the oxide product was a fine powder. At higher temperatures (973 K to 1173 K) oxidation slowed due to a densification process with formation of coarse fragments. The oxide fragmentation at 973 K and 1073 K and oxide pulverisation at 873 K were observed in situ in a HT-ESEM. Cracking induced fragmentation and pulverisation was linked to stresses generated from the volumetric transformation from UC to U 3 O 8. [ABSTRACT FROM AUTHOR]

Highlights • Thermal analysis of alternative ceramic fuels in the HPLWR fuel assembly using CFD codes was carried out. • Thermodynamic advantages of UN and UC make them promising as nuclear fuels in SCWRs. • The centerline temperatures of UN-ZrC and UC-TiN fuels, were lower than those obtained with UO 2 • The UN-ZrC and UC-TiN fuels showed a more flattened radial temperature profile than UO 2. • The maximum temperature of UN-ZrC and UC-TiN fuels has a minimal dependence on the manufactured porosity. Abstract The High Performance Light Water Reactor (HPLWR) is one of the most promising designs of the Super-Critical Water Reactor (SCWR). The use of uranium nitride (UN) and uranium carbide (UC), as alternative nuclear fuels for the SCWR, offer the advantage of high thermal conductivity compared to uranium dioxide (UO 2). For the analysis of these alternative nuclear fuels in SCWRs, some important design features must be considered. One of them is the porosity; the nuclear fuel is manufactured with different porosity values, as high as 20%, to reduce its hardness and swelling, influencing its thermal conductivity. Another issue is related to the chemical reactivity of UN and UC with water and nickel, which forces the use of coatings for fuel pellets. In this paper, a preliminary thermal-hydraulic analysis, using CFD codes, of the HPLWR fuel assembly, focused on the thermal behavior of ceramic fuel, was carried out. The use of UN coated with zirconium carbide layers and UC coated with titanium nitride layers were analyzed. The impact of porosity induced changes in fuel temperature profiles was analyzed. The radial and axial fuel temperature distributions were obtained for all cases. It was found that the maximum temperature values obtained using UN and UC, both coated and uncoated, were lower than those obtained with UO 2. It was found that the porosity changes on the proposed fuels have a small influence on the maximum fuel temperature. [ABSTRACT FROM AUTHOR]

Gas-phase conversion of UC in NOx-air, NOx-H20 (vapor)-air, and HN03 (vapor)-air atmospheres in the temperature interval from 298 to 673 K was studied. In the course of the gas-phase conversion in an NOx-air atmosphere, UC transforms into uranium oxides and oxyhydroxides; uranyl nitrates and hydroxyni-trates are formed in smaller amounts. In NOx-H20 (vapor)-air and HN03 (vapor)-air atmospheres, the major conversion products are uranyl nitrate hydrates, with the nitration occurring at a lower temperature than in an NOx-air atmosphere. [ABSTRACT FROM AUTHOR]

The studies of the synthesis of uranium carbide from uranyl-ascorbate gels using the complex sol-gel process (CSGP) have been carried out. The synthesis of uranyl-ascorbate mixture as liquid sol from uranium trioxide and ascorbic acid and solid gel by extraction of water from sol were carefully examined. Ascorbic acid was used as a complexing agent in complex sol-gel process and as a carbon source. The crucial step to obtain final uranium carbides from the aforementioned substrates is the carbonization process. The thermal behavior of ascorbic acid and uranyl-ascorbate gels in a nitrogen atmosphere in the temperature range of 25-900°C were investigated using TG-DTG. Furthermore, the products of the carbonization of uranyl-ascorbate gels in nitrogen, argon and vacuum atmosphere were identified by X-ray diffraction. TG-DTG was used also as a method for determining of carbon residues in the samples.

Abstract The optimization of the microstructure of the UC x target is a key point since many years in the field of ISOL method. The ultimate goal is to facilitate the release of the fission products, especially those with short half-lives. Fourteen UC x samples were synthetized from different uranium and carbon sources using three mixing protocols. All carburized samples were systematically characterized in terms of nature and proportion phases, grain and aggregate size, open and close porosity proportion and open pore size distribution. Our results were analysed using a multivariate statistical approach in order to remove any subjective bias. Strong correlations between the physicochemical characteristics of the samples as well as the impact of the synthesis process have been highlighted. In particular, using carbon nanotubes as carbon source combined with a new method of mixing is the key parameter to limit the sintering and to obtain samples with small grains and a high porosity well distributed over small pores. Moreover the microstructure obtained proved to be stable at high temperature. [ABSTRACT FROM AUTHOR]

[Display omitted] • Migration of microparticle impurities in the uranium melt inside an induction furnace. • Homogeneous distribution of smaller carbide microparticles (less than50 μm) is predicted. • Larger particles (>50 μm) tend to move toward the crucible wall and upward in the system. • Particle shape does not have a significant impact on particle distribution in the melt. Understanding and controlling impurity behavior in actinide metal casting processes are foundational for efficient part production yet remain major challenges for researchers and industry. To help provide insight regarding impurity distribution during actinide metal casting, we have developed computational fluid dynamics (CFD) models for a laboratory-scale system using commercial and open-source codes. This article describes multiple experiment-informed models designed to simulate a specific laboratory system in which uranium melt, containing a known starting concentration of carbon impurity, is electromagnetically stirred in an induction furnace. The goal of the simulations is to predict the motion of impurity particles in the form of uranium monocarbide in the velocity field of the melt. Prior to simulating the uranium–carbon system, numerical models were validated using a previously published nonradioactive experimental system. Effects of the size and shape of impurity particles in the models were investigated and agree with experimental findings. Simulation of smaller particles (less than 50 μm) shows more homogenous distribution throughout the stirred melt. With increased particle size (100 μm), the body forces, which include the buoyancy force, dominate over the drag force, causing larger particles to move toward the crucible walls and upward in the system. [ABSTRACT FROM AUTHOR]

The Energy Multiplier Module (EM2) is a helium-cooled fast reactor with a core outlet temperature of 850°C. It is designed as a modular, grid-capable power source with a net unit output of 265 MWe. The reactor employs a convert-and-burn core design that converts fertile isotopes to fissile and burns them in situ over a 30-year core life. The reactor is sited in a below-grade sealed containment. It uses passive safety methods for heat removal and reactivity control to protect the integrity of the fuel, reactor vessel, and containment. The plant also incorporates a below-grade, passively cooled spent fuel storage facility with capacity for 60 years of full-power operation. EM2employs a direct closed-cycle gas turbine power conversion unit (PCU) with an organic Rankine bottoming cycle for 53% net power conversion efficiency assuming evaporative cooling. The high-power conversion efficiency and long-burn fuel cycle reduce the electricity cost by 35% when compared with the conventional light water reactor. The conceptual design has been conducted for the EM2plant with focus on the reactor, fuel, and safety system designs. A detailed model of the passive direct reactor auxiliary cooling system was created to demonstrate functionality for selected design-basis accidents. The bench-scale fuel development campaign demonstrated high-quality uranium carbide pellet fabrication as well as β-SiC composite cladding and SiC-joining technologies. Irradiation tests of reactor materials are also being conducted. The PCU variable-speed generator mechanical design was validated with operational testing of its novel rotor at speeds >13 000 rpm. The design of the turbo-compressor with active magnetic bearings continues. A large cost database and financial model have been constructed for use as a key driver for the design to be economically competitive with competing generating technologies after 2030. [ABSTRACT FROM PUBLISHER]

In situ HT-ESEM oxidation of sintered UC fragments revealed the morphological changes occurring during the transformation between UC to UO 2 and UO 2 to U 3 O 8 at 723–848 K and in an atmosphere of 10–100 Pa O 2 . Two main oxidation pathways were revealed. Oxidation at 723 K in atmospheres ≤25 Pa O 2 showed the transformation from UC to UO 2+x , as confirmed by post mortem HRTEM analysis. This oxidation pathway was comprised of three steps: (i) an induction period, where only surface UC particles oxidised, (ii) a sample area expansion accompanied by crack formation and propagation, (iii) a stabilisation of the total crack length inferring that crack propagation had stopped. Samples oxidised under 50 Pa O 2 at 723 K and at 773–848 K for 10–100 Pa O 2 showed an “explosive” oxidation pathway: (i) sample area expansion occurred as soon as oxygen was inserted into the chamber and crack propagation and crack length followed an exponential law; (ii) cracks propagated as a network and the oxide layer fragmented, (iii) an “explosion” occurred causing a popcorn-like transformation, typical for oxidation from UO 2 to U 3 O 8 . HRTEM characterisation revealed U 3 O 8 preferentially grow in the [001] direction. The explosive growth, triggered by ignition of UC, proceeded as a self-propagating high-temperature synthesis reaction, with a propagation speed of 150–500 ± 50 μm/s. [ABSTRACT FROM AUTHOR]

Fully Ceramic Micro-Encapsulated (FCM) fuel has greater capability of fission product retention and oxidation resistance. Presence of an additional fission product barrier in the form of strong SiC layer, make it an attractive option to be used as PWR fuel. In present studies, PWR fuel assemblies are transformed from standard 17 × 17 into new designs of 15 × 15 and 13 × 13 arrays with increased rod diameter but the same assembly dimension. These assemblies contain TRISO type fuel embedded in SiC matrix. For full implementation of the accident tolerant fuel concept, this study also investigates on the replacement of conventional Zircaloy-4 cladding with SS-304 and FeCrAl to minimize the hydrogen production due to interaction of water with cladding at high temperatures. Both are found to be feasible as a clad, however the neutronic penalty associated with SS-304 is very high, ultimately leading to shorter cycle length. A preliminary study of burnable poisons i.e. Erbia and Gadolinia as QUADRISO particles shows that the Erbia is preferable material due to its good depletion performance. In new assembly designs, temperature coefficients of reactivity (Doppler & Moderator) are negative through the life cycle of FCM fuel. Pin power peaking factors are less than 1.10 at BOC and RCCA worth is higher compared with reference assembly. Transuranic production is almost two to three times lower, compared with reference UO 2 fuel assembly. This is very important aspect of FCM fuel from proliferation point and viewed as an additional advantage. [ABSTRACT FROM AUTHOR]

Uranium carbide is a candidate fuel for Generation IV nuclear reactors. However, like any candidate fuel, a reprocessing route should be established before implementation. One proposed method involves a pre-oxidation step, where the carbide fuel is oxidized to an oxide and then reprocessed as normal. A mathematical model has been developed to simulate such an oxidation using finite difference approximations of the heat and mass transfer processes occurring. Available literature was consulted to provide coefficients for the reaction rates and importantly the diffusion of oxygen through the adherent oxide layer that forms on the carbide: the rate limiting step. The transient temperature, oxygen and carbon monoxide distributions through the system are modeled in order to predict oxidation completion times and the temperatures reached. It was found that for a spherical pellet of radius 0.935 cm, the oxidation can take from 1 to 19 h depending on the oxidation conditions and reach temperatures of up to 1556 °C. A robust model results that offers increased understanding of a process crucial to the sustainable use of carbide fuels in energy generation. [ABSTRACT FROM AUTHOR]

• UC possesses a locally disordered structure revealed by EXAFS. • The onset temperature of UC bulk oxidation was determined to be 537 K in air. • UC follows a complex stepwise thermal oxidation process in air largely driven by the presence of carbon. • Standard enthalpy of formation of UC was determined by high temperature solution drop calorimetry at 973 K using a TiO 2 pellet encapsulation method. • Standard enthalpies of formation of UC, U 2 C 3 , and α-UC 1.94 exhibit a negative linear correlation with the C/U molar ratio. To enable better implementation of uranium monocarbide (UC) as an advanced nuclear fuel for future high-temperature reactors, it is essential to have a thorough knowledge of its thermal and thermodynamic properties under reactor operational conditions. In this work, we studied thermal bulk oxidation of UC by simultaneous thermal analysis consisting of thermogravimetric analysis – differential scanning calorimetry coupled with evolved gas mass spectrometry (TGA-DSC-MS), and we examined the thermodynamic stability of UC using high temperature oxide melt drop solution calorimetry. In air, our studied UC sample (which contains ∼5 mol% UO 2) was found to undergo a step-wise thermal oxidation process consisting of consecutive oxidations and thermal decomposition reactions: 0.95UC·0.05UO 2 → UO 3 ·0.29(C x O y) + 0.66CO 2 → UO 3 ·0.20(C x O y) + 0.09CO 2 → UO 3 ·0.03(C x O y) + 0.17CO 2 → U 3 O 8 + 0.03CO 2 + 0.166O 2. DSC was further used to determine the enthalpies of reactions associated with this series of oxidation reactions. Synchrotron X-ray diffraction (XRD) and extended X-ray absorption spectroscopy (EXAFS) were performed to characterize both the long- and short-range structures of UC. The standard enthalpy of formation (Δ H ° f) of UC was determined to be –50.7 ± 10.8 kJ/mol·atom, in good agreement with previous values measured by bomb calorimetry. Lastly, the enthalpic landscape of U-C compounds, including UC, U 2 C 3 , and α-UC 1.94 , were established based on the enthalpy normalized per mole atom, which suggests that U-C phases are thermodynamically stable at lower C/U ratios. [ABSTRACT FROM AUTHOR]

Results of uranium carbide sintering using a spark plasma sintering facility are presented here. The initial uranium carbide powder is produced from the thermal decomposition of a citric acid and uranium nitrate mixture. The study shows that spark plasma sintering is a very efficient compaction tool for uranium carbide material. Both onset of sintering and final density are strongly correlated to the UC synthesis conditions. [ABSTRACT FROM AUTHOR]

The preparation of uranium carbide (UC) by carbothermal reduction and its sintering into dense pellets by conventional means require high temperatures for long periods. We have developed a preparation route yielding fine UC powder with significantly increased sinterability. At first, a mixture of nanocrystalline UO 2 embedded in amorphous carbon (nano-UO 2 /C) was obtained by thermal decomposition of a gel containing solubilised uranyl nitrate and citric acid. Later, the nano-UO 2 /C powder was treated in a conventional furnace or in a modified spark plasma sintering facility at elevated temperatures (≥1200 °C) in order to obtain uranium carbide powder. The effects of initial composition, temperature, gas/vacuum atmosphere and the overall reaction kinetics are reported. [ABSTRACT FROM AUTHOR]

A novel technique to fabricate precursors of carbide-ceramic nuclear fuel via synthesis of spherical particles (with a diameter below 150 μm) of carbon black dispersion in uranyl-ascorbate gels by a combination of the Complex Sol–Gel Process and the Double Extraction Process has been elaborated. Applying of ascorbic acid as complexing agent in the Complex Sol–Gel Process allows to obtain stable and uniform dispersion of carbon in an uranyl-ascorbate sol. Traditionally, the dispersability is improved by employment of dispersing agents – surfactants. In the presented Complex Sol–Gel Process, a dispersing agent is needless. The ascorbate sols perfectly wet carbon black and do not form agglomerates. In the Double Extraction Process, dispersion of carbon black in sols is emulsified by drops in the organic phase consisting of 2-ethylhexanol-1 and 1 volume % of Primene JM-T. The microspheres are created by gelation of droplets, through simultaneous extraction of water with 2-ethylhexanol-1 and nitrate ions with Primene JM-T. The elaboration of the optimal gelation conditions requires considering many issues which are described in this paper. [ABSTRACT FROM AUTHOR]

In the framework of a R&D program aiming to develop uranium carbide (UC x ) targets for radioactive nuclear beams, the Institut de Physique Nucléaire d’Orsay (IPNO) has developed an experimental setup to characterize the release of various fission fragments from UC x samples at high temperature. The results obtained in a previous study have demonstrated the feasibility of the method and started to correlate the structural properties of the samples and their behavior in terms of nuclear reaction product release. In the present study, seven UC x samples have been systematically characterized in order to better understand the correlation between their physicochemical characteristics and release properties. Two very different samples, the first one composed of dense UC and the second one of highly porous UC x made of multi-wall carbon nanotubes, were provided by the ActILab (ENSAR) collaboration. The others were synthesized at IPNO. The systems for irradiation and heating necessary for the release studies have been improved with respect to those used in previous studies. The results show that the open porosity is hardly the limiting factor for the fission product release. The homogeneity of the microstructure and the pore size distribution contributes significantly to the increase of the release. The use of carbon nanotubes in place of traditional micrometric graphite particles appears to be promising, even if the homogeneity of the microstructure can still be enhanced. [ABSTRACT FROM AUTHOR]

Uranium carbide is a candidate fuel for future nuclear reactors. However, for it to be implemented in a closed fuel cycle, an outline for its reprocessing is necessary. One proposed method is to oxidise the uranium carbide into uranium oxide which can then be reprocessed using current infrastructure. A mathematical model describing the heat and mass transfer processes involved in such an oxidation has been constructed. The available literature was consulted for reaction coefficients and information on reaction products. A stable and convergent numerical solution has been developed using a combination of finite-difference approximations of the differential equations. Completion times of approximately 3–30 h are predicted given a spherical pellet with a radius of 9.35 mm under varying initial conditions. The transient temperature distribution throughout the system is predicted, with a maximum temperature of 1458 °C observed from an initial temperature of 500 °C at an oxygen concentration of 3.15 mol m −3 . [ABSTRACT FROM AUTHOR]

The measurements of actinide isotopic ratios, in particular 236 U/ 238 U, in environmental samples requires high sensitivity. In particular, special effort must be devoted to the suppression of interfering nuclides, such as 235,238 U, when measuring 236 U. At the AMS facility of CIRCE, isotopic ratios down to ∼10 −10 are currently measured using a gas E − Δ E detector. In order to push this limit lower towards natural levels, a time-of-flight system is used, featuring a micro-channel plate start detector and a Si stop detector. As the mass resolution of the latter is limited by the layout, an attempt to reduce the abundant isotope interference by other means has been undertaken. In this study, we report preliminary results on the characterization of the presence of molecular interferences when using UO − , UC − and UC 2 − as injected molecular ions. In particular the possibility to stabilize the current yield from carbide cathodes has been investigated: it was found that the best cathode preparation procedure is obtained pressing U salts baked at 800 °C mixed with graphite and Al powder. Moreover, the 238 U background in 236 U measurements is strongly reduced injecting UC − ions, as verified using a time-of-flight detection system. On the other hand, 235 U interference is larger with respect to UO − injection, but this increase is smaller than expected on the basis of abundances of 13 C and 17 O in UC and UC 2 molecules on one side and UO, on the other. [ABSTRACT FROM AUTHOR]

Thermal conductivity and emissivity measurements on different types of uranium carbide are presented, in the context of the ActiLab Work Package in ENSAR, a project within the 7th Framework Program of the European Commission. Two specific techniques were used to carry out the measurements, both taking place in a laboratory dedicated to the research and development of materials for the SPES (Selective Production of Exotic Species) target. In the case of thermal conductivity, estimation of the dependence of this property on temperature was obtained using the inverse parameter estimation method, taking as a reference temperature and emissivity measurements. Emissivity at different temperatures was obtained for several types of uranium carbide using a dual frequency infrared pyrometer. Differences between the analyzed materials are discussed according to their compositional and microstructural properties. The obtainment of this type of information can help to carefully design materials to be capable of working under extreme conditions in next-generation ISOL (Isotope Separation On-Line) facilities for the generation of radioactive ion beams. [ABSTRACT FROM AUTHOR]

Uranium and plutonium carbides are candidate fuels for Generation IV nuclear reactors. This study is focused on the characterization of uranium monocarbide samples. The successive fabrication steps were carried out under atmospheres containing low oxygen and moisture concentrations (typically less than 100 ppm) but sample transfers occurred in air. Six samples were sliced from four pellets elaborated by carbothermic reaction under vacuum. Little presence of UC 2 is expected in these samples. The α -UC 2 phase was indeed detected within one of these UC samples during an XRD experiment performed with synchrotron radiation. Moreover, oxygen content at the surface of these samples was depth profiled using a recently developed nuclear reaction analysis method. Large oxygen concentrations were measured in the first micron below the sample surface and particularly in the first 100–150 nm. UC 2 inclusions were found to be more oxidized than the surrounding matrix. This work points out to the fact that more care must be given at each step of UC fabrication since the material readily reacts with oxygen and moisture. A new glovebox facility using a highly purified atmosphere is currently being built in order to obtain single phase UC samples of better purity. [ABSTRACT FROM AUTHOR]

The direct dissolution of UC, a potential fuel for generation-IV nuclear reactors, in nitric acid can produce organic complexants that may interfere with the consumed fuel reprocessing based on liquid-liquid extraction. This study shows the presence of 17 % of carboxylic and polycarboxylic acids (e.g. benzenepentacarboxylic acid). 25 % of the carbon mass balance seemed to correspond to unfunctionalized polyaromatic compounds with more than 3-4 aromatic rings. Under an active oxidation process, these compounds become smaller and very functionalized polyaromatic compounds. Structures have been proposed for 8 % of the carbon mass balance. [ABSTRACT FROM AUTHOR]