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Uranium carbide (UC) is a candidate fuel material for future Generation IV nuclear reactors. As part of a general safety assessment, it is important to understand how fuel materials behave in aqueous systems in the event of accidents or upon complete barrier failure in a geological repository for spent nuclear fuel. As irradiated nuclear fuel is radioactive, it is important to consider radiolysis of water as a process where strongly oxidizing species can be produced. These species may display high reactivity toward the fuel itself and thereby influence its integrity. The most important radiolytic oxidant under repository conditions has been shown to be H2O2. In this work, we have studied the dissolution of uranium upon exposure of UC powder to aqueous solutions containing HCO3- and H2O2, separately and in combination. The experiments show that UC dissolves quite readily in aqueous solution containing 10 mM HCO3- and that the presence of H2O2 increases the dissolution further. UC also dissolves in pure water after the addition of H2O2, but more slowly than in solutions containing both HCO3- and H2O2. The experimental results are discussed in view of possible mechanisms.

International audience; The optimization of the microstructure of the UCx 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 UCx 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.

During the production of uranium oxide microspheres with carbon using internal gelation, a small fraction of microspheres will not meet the required size or sphericity specification. The uranium microspheres with carbon can be rejected after they have been air-dried or converted into uranium carbide and uranium oxide (UCO) kernels. The next step for the rejected spheres was an air oxidation for carbon removal. The air-dried spheres became triuranium octoxide (U3O8) spheres, which were sometimes ground into powder. The UCO kernels became U3O8 powder during the air oxidation. The next recycle step was nitric acid dissolution of the U3O8 spheres or powders to produce acid-deficient uranyl nitrate (ADUN) solutions, which were used to make new uranium oxide microspheres with carbon and subsequently UCO kernels. The kinetics of the acid dissolution process were compared, and the impurity levels in the different ADUN solutions were determined. X-ray diffraction results for the UCO kernels from the initial and recycled ADUN solutions indicate that changes in the impurity levels can impact the uranium carbide to uranium dicarbide ratio in the UCO kernels.

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 ...

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO2-kernel TRISO particles have undergone 1600 °C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes (110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr) were monitored while heating the fuel specimens to 1600 °C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600 °C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600 °C. However, additional silver release was observed later in the safety testing due to the UO2 TRISO with failed SiC. Failure of the SiC layer in the UO2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.

Uranium carbides are receiving renewed interest as a preferred nuclear fuel composition for advanced reactors due to their numerous favorable properties. Like many other refractory and transition metal carbides these compounds exist in both hypo- and hyper-stoichiometric compositions which results in significant variations in physical, thermal, and mechanical properties. This manuscript surveys both historic and recent literature to compile important properties data mainly as a function of temperature and carbon content to support fuel performance modeling. Expressions and fits for such properties are also suggested when allowed by complete data sets. The manuscript also attempts to highlight gaps and discrepancies in the reported data to assist the nuclear fuels community identify key areas that may require further attention in terms of future research and development.

Uranium silicide ( U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel which operates at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. To support the implementation of a U 3 Si 2 fission gas model in BISON, cluster dynamics simulations of irradiation enhanced Xe diffusion have been carried out. Similarly, MD simulations were used to predict the athermal contribution due to atomic mixing during ballistic damage cascades. By combining our results with literature DFT data for thermal equilibrium diffusion, Xe diffusivity has been described over a wide range of temperatures for in-reactor conditions. These lower length scale informed models are then utilized in the assessment of BISON U 3 Si 2 modeling capabilities by simulating the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the assessment process to identify where further experiments and lower length scale modeling would be beneficial. The multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

Human need of energy will increase time to time. Therefore, a safe, renewable, and efficient source of energy, which is Nuclear Energy, is needed. Nuclear Power Plant (NPP) is the most compatible solution to provide electricity to human race in the future. The problem that came within NPP is the danger of proliferation issues. The method that has been developed to overcome this problem is CANDLE [5] and has been modified by Prof. Zaki Su’ud (Modified CANDLE scheme). This research use Axial Modified CANDLE Scheme to Helium-Cooled Fast Reactor with Natural Uranium Carbide-Thorium Carbide as fuel and applied to various size of core as optimization. Neutronic aspect such as, burn up level, multiplication factor, and conversion ratio are utilized in this paper in order to analyse the behaviour of the reactor. Other than that, percentage of Uranium has been varied to reduce power peaking. The neutronic calculation has been done using SRAC and core design calculation by FI-ITB-CH1. This research concludes that power peaking reduction is able to achieve by combining Uranium Carbide and Thorium Carbide to the fuel. The optimum reactor design reached at 360 cm of core radius and 303 cm of core height.

In this paper, we investigated the neutronics performance of uranium carbide (UC), uranium nitride (UN) and uranium dioxide (UO2) monolith targets and granular targets by using GPU-based Monte Carlo particle transport program (GMT). Effects of the target thickness and diameter on leaked neutron yield of monolith targets irradiated by 1 GeV protons were studied; the leaked neutron yield and neutron spectra for the dense granular uranium compounds and UC-kernel TRISO type targets were presented. Results show that, with an optimal dimension ϕ 25 × 50 cm , uranium compounds target can produce 40% more leaked neutron yield than the tungsten target. Results also show that by 1 GeV protons incident on, our designed carbon and SiC layers coated UC-kernel TRISO type pellets target with about 70% kernel volume fraction can produce about 30% more leaked neutron yield than the granular tungsten target in their optimal diameters.

The dissolution of uranium or uranium-plutonium carbide fuel in nitric acid leads to ~50% carbon evolved as carbon dioxide, the remainder remains in the solution as soluble organics. These dissolved organic molecules interfere with the solvent extraction of uranium and plutonium by complexing to the actinide ions and decreasing the efficiency of their extraction. Experiments reported here describe two series of experiments assessing the uranium carbide dissolution liquor treatment by prolonged boiling and electrochemical oxidation. Plutonium losses to aqueous and solvent raffinates are observed for untreated liquors, highlighting that mineralisation of dissolved organics is necessary to reduce the complexing effects of organic acids to an extent that permit efficient operation of a solvent extraction process both in the first solvent use (considered here) and for maintaining solvent quality during industrial solvent reuse in the highly active cycle. Solution carbon analysis and 30% TBP solvent extraction batch tests of uranium and plutonium originating from dissolved uranium carbide liquors untreated and after treatment are compared. These experiments demonstrate the reprocessing of uranium carbides by direct dissolution coupled to a mineralisation process, can achieve near quantitative uranium and high plutonium recoveries (99.9%).

Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

Producción Científica, Uranium compounds are used as fissile materials in nuclear reactors. In present day reactors the most used nuclear fuel is uranium dioxide, but in generation-IV reactors other compounds are also being considered, such as uranium carbide and uranium mononitride. Upon possible accidents where the coolant would not circulate or be lost the core of the reactor would reach very high temperatures, and therefore it is essential to understand the behaviour of the nuclear fuel under such conditions for proper risk assessment. We consider here molten metallic uranium at several temperatures ranging from 1455 to 2050 K. Even though metallic uranium is not a candidate for nuclear fuel it could nevertheless be produced due to the thermochemical instability of uranium nitride at high temperatures. We use first principles techniques to analyse the behaviour of this system and obtain basic structural and dynamic properties, as well as some thermodynamic and transport properties, including atomic diffusion and viscosity., Junta de Castilla y León (Ref. project VA124G18), Ministerio de Economía, Industria y Competitividad (Project PGC2018-093745-B-I00) and FEDER

To study the influence of the structural properties of UCx 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 {\mu}A, 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., Comment: 11 pages, 5 figures, 1 table

International audience; 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 U3O8 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 U3O8.

Survey information on integral material studies of advanced fuel compositions performed at the Institute of Reactor Materials over a period of 50 years is presented. The influence of the reactor testing conditions on the radiation resistance of different forms of nuclear fuel was investigated: oxide, carbide, nitride, carbosulfide, and some of their compounds. Uranium oxides possessed a different initial microstructure and were doped with the oxides Y2O3, TiO2, Sc2O3, and others. Carbide fuel was tested in the form of solid solutions of uranium carbide with Zr, Nb, and Ta carbides. The dimensional, phase, and structural stability of fuel compositions were studied in hot boxes; the electrophysical and mechanical properties, such as strength in compression and bending, Young’s modulus, the shear modulus, and Poisson’s ratio were investigated. The experimental data on the radiation resistance of advanced fuel compositions formed the basis for the validation of the serviceability of different nuclear reactors.

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release

The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610°C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr2O3 , and a low amount of uranium metal. After annealing at 610°C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

2nd International Conference on Computational and Experimental Science and Engineering (ICCESEN) -- OCT 14-19, 2015 -- Kemer, TURKEY, WOS: 000384810700018, In this study, the medical radioisotope production performance of a conceptual accelerator-driven system is investigated. Lead-bismuth eutectic is used as target material. The fuel core of the considered accelerator-driven system is divided into ten subzones, loaded with uranium carbide and various isotopes (isotopes of copper, gold, cobalt, holmium, rhenium, scandium, and thulium) and cooled with light water. As is known, light water is an effective moderator of neutrons as well as a good coolant. The fuel and the isotopes are separately placed as cylindrical rods with a cladding of carbon composite. The volume fractions of fuel, isotope, cladding and coolant are selected as 25%, 35%, 10% and 30%, respectively. The copper rods are placed into the first five subzones due to the fact that copper isotopes have low capture cross-section. In the case of the each radioisotope production, one of the other considered isotopes that have higher capture cross-section are placed into the following five subzones for optimization of fission, fissile breeding and radioisotope production. The graphite zone is located around the fuel core to reflect the escaping neutrons. Boron carbide (B4C) is used as shielding material. In order to produce more neutrons (about 25-30 neutrons per 1 GeV proton), the target is irradiated with a continuous beam of 1 GeV protons. All neutronic computations have been performed with the high-energy Monte Carlo N-Particle Transport Code using the LA150 data library. The neutronic results obtained from these calculations show that the examined accelerator-driven system has a high neutronic capability, in terms of production of thermal power, fissile fuels, and medical radioisotopes.

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.

We investigated the formation of UCx from UO2+x and graphite in situ using neutron diffraction at high temperatures with particular focus on resolving the conflicting reports on the crystal structure of non-quenchable cubic UC2. The agents were UO2 nanopowder, which closely imitates nano grains observed in spent reactor fuels, and graphite powder. In situ neutron diffraction revealed the onset of the UO2 + 2C → UC + CO2 reaction at 1440 °C, with its completion at 1500 °C. Upon further heating, carbon diffuses into the uranium carbide forming C2 groups at the octahedral sites. This resulting high temperature cubic UC2 phase is similar to the NaCl-type structure as proposed by Bowman et al. Our novel experimental data provide insights into the mechanism and kinetics of formation of UC as well as characteristics of the high temperature cubic UC2 phase which agree with proposed rotational rehybridization found from simulations by Wen et al.

Uranium carbides have attracted renewed interest as advanced nuclear fuels for Generation IV reactors. As an important property required for gas bubble modeling in nuclear fuels, the surface energy of uranium carbides is scarce in literature. In this work, we study the surface properties of uranium carbides by first-principles density functional theory calculations. Surface orientations with maximum Miller index up to 3, 2 and 2 are investigated for UC, U2C3 and α-UC2, respectively. By studying the effects of surface termination and chemical potential on surface energy, we identify the factors that determines the surface stability. From the calculated surface energies, the surface properties of uranium carbide single crystals are obtained from Wulff construction, including equilibrium morphology, dominant surface orientation and area weighted surface energy.

Ceramic nuclear fuels may have potential benefits to be considered as nuclear fuels in future Supercritical Water Reactors (SCWR), they offer the advantage of high thermal conductivity and high-density values compared to uranium dioxide. In recent years, uranium nitride (UN) and uranium carbide (UC) are being considered as viable options for alternative nuclear fuels in SCWR, but an important issue is related to their chemical reactivity with water and nickel, which forces the use of coatings for fuel pellets, reducing heavy metal volume and reducing their thermal and neutronic performance. The main purpose of the study is the analysis of the burnup dependent performance of coated ceramic fuels in the SCWR compared with conventional UO2 fuel. Particularly, it focuses on the study of the fuel centerline temperature and the cladding surface temperature. Furthermore, the evolution of the neutron multiplication factor, the fissile inventory ratios and the main heavy isotopes composition along burnup are analyzed and compared to asses proliferation resistance issues. The use of UN fuel coated with Zirconium Carbide (ZrC) and UC fuel coated with Titanium Nitride (TiN) in the fuel assembly of the High-Performance Light Water Reactor is analyzed, and the results are compared with those of the conventional UO2 fuel. The comparative study was conducted for a hypothetic operation time of 540 days. Due to the large coolant density variation along with the active height of the fuel assembly, coupled calculations are performed using the Monte Carlo MCNP6 code and the CFD code ANSYS-CFX 19, taking into account the capabilities of CFD codes to describe the heat transfer mechanisms at supercritical conditions. The main results show significantly lower values of the centerline temperature distribution in UN_ZrC and UC_TiN fuels compared to UO2 fuel. The fuel burnup values obtained at the end of the 540 days are practically the same, a slightly higher burnup is obtained for UC_TiN fuel because of its lower amount of heavy metals in the initial composition. The evolution of the fissile inventory is practically the same for all the analyzed fuel options, around 65% of fissile inventory is still present in the core at EoC. Total plutonium production shows a decrease of 1.85% (UN–ZrC) and 4.53% (UC–TiN) with respect to UO2 fuel. Coated ceramic fuels show a slightly improved proliferation resistance compared with UO2 fuel, since a slightly smaller fraction of 239Pu is obtained in their compositions.

In the Sol–Gel Laboratory at the Institute of Nuclear Chemistry and Technology (INCT), studies of synthesis of various advanced ceramic materials by the sol–gel method have been carried out for 50 years. The ceramics products, such as metal oxides, homogeneous mixtures of metal oxides, cermetals and metals, were obtained in various shapes, as irregular powders, monoliths, coatings, fibers or spherical particles with various diameters. A new method using a water-in-oil technique – the ICHTJ Process (INCT in English) was elaborated to synthesize spherical particles of metal oxide or mixture of metal oxides, with diameters below 100 µm. Through years, the method has been evaluated and optimized. Prepared sol solutions of metal cations are emulsified in 2-ethylhexanol-1 containing 1 volume % of the surfactant sorbitol monooleate. Drops of emulsion are gelled by extracting their water with this organic solvent. The main parameters e.g. flow rate; gelation time, syringe, sol solution etc. that have a strong influence on the optimization of the synthesis method and features of the spherical particles are described in this article through an example of synthesis of uranium oxycarbide – mixture of uranium dioxide with uranium carbide.

We use imaging secondary ion mass spectrometry (SIMS) with a Cameca NanoSIMS 50 to determine the distribution and relative concentration of H in two uranium metal samples: sample1(LANL), as cast with less than 1 wppm H, sample2(Y12), outgassed 6 h in vacuum at 630 °C to remove hydrogen. H ion counts appear almost exclusively associated with ‘carbide’ inclusions, based on H − , C − and O − ion images for uranium surfaces sputter-cleaned in situ with a 16 keV Cs + ion beam. Two classes of inclusions are identified: small, micrometer to sub-micrometer inclusions and larger, clearly angular inclusion (⩾3 μm). In sample1(LANL) the large inclusions (⩾250/mm 2 ) show a low H − / C − ratio inside, and have H − / C − ratios at their perimeters comparable in magnitude to that seen in μm-size inclusions. Small inclusions (∼2500/mm 2 ) contain H more uniformly throughout and, averaged over the inclusion, the small inclusions have approximately 50 times higher relative H concentration than the large inclusions. Sample2(Y12) was found to have comparable H − / C − ratios in the large carbides, but no small inclusions were observed. Because of the matrix-sensitivity of SIMS, H / C ratios representing the actual composition of the inclusions cannot be derived from the H − / C − ratios without calibration UC samples with known H content, which are not currently available.

We investigate the electronic thermal conductivity of alternative fuels like uranium nitride and uranium carbide. We evaluate the electronic contribution to the thermal conductivity, by combining first-principles quantum-mechanical calculations with semiclassical correlations. The electronic structure of UN and UC was calculated using Quantum Espresso code. The spin polarized calculations were performed for a ferromagnetic and antiferromagnetic ordering of magnetic moments on uranium lattice and magnetic moment in UC was lower than in UN due to stronger hybridization between 2p electrons of carbon and 5f electrons of uranium. The nonmagnetic electronic structure calculations were used as an input to BolzTrap code that was used to evaluate the electronic thermal conductivity. It is predicted that the thermal conductivity should increase with the temperature increase, but to get a quantitative agreement with the experiment at higher temperatures the interaction of electrons with phonons (and electron-electron scattering) needs to be included.

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.

In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 °C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 × 10−2 for silver, 1.4 × 10−3 for europium, and 1.1 × 10−5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 °C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X-ray imaging of selected particles was performed to examine the internal microstructure. This examination revealed variable irradiation performance of the coating layers, but sufficient statistical sampling is not yet available to identify any possible correlation to variation in individual particle fission product retention.

The properties of the high-temperature monocompounds of uranium UO2, UN, UC, US, and UP are analyzed for the operating conditions of the fuel elements in thermionic nuclear power installations. The advantages and disadvantages of each of these fuels, whose properties do not meet entirely the specifications for long-life thermionic fuel elements, are shown. The results obtained at NII NPO Luch on the basis of these monocompounds of uranium are presented: sub-stoichiometric uranium dioxide with thermally stabilized, predominantly open porosity and with columnar optimized grain structure as well as high-thermalconductivity fuel compositions – solid solutions based on uranium carbide UC–TaC(ZrC), uraniumzirconium carbonitride and uranium carbosulfide. The uses best suited for each of these fuels are determined. A promising design for fuel elements with a multirod kernel, making it possible to solve the problem of removing gaseous fission products from the fuel-element cavity and securing the geometric stability of cladding, is presented for the fuel compositions UC–TaC(ZrC), whose exhibit the lowest evaporation rate.

Defects such as the most stable hexavacancy (V6) distribute widely on neutron-irradiated graphite surfaces, which play a dominant role in immobilizing radioactive products released from nuclear fuels. By performing DFT calculations, we explore the interaction of gaseous uranium dicarbide (UC2) molecules on a graphene nanosheet with a V6 defect, in order to investigate the behavior of the representative vapor species of uranium carbide fuels in reactor cores. Results suggest that UC2 can be trapped in the V6 defect with considerable binding energy of >10 eV, with all the six dangling bonds of the V6 defect being saturated by UC2. Bonding nature analyses also reveal that the U–C interaction lies in the synergistic interplay between electrostatic and covalent interaction with extensive participation of U valence electrons from 5f to 7p orbitals, which further stimulate polarization of semi-core 6p orbitals and their subsequent contributions to the bonding. This strong interaction leads to a favorable binding of UC2 to the defective graphite surface, which reduces the capability of nuclear graphite to retain harmful fission products by the vacancies being filled with UC2. These findings highlight substantial chemical reactivity and strong localization of UC2 on the widespread V6 defects in nuclear graphite, and may provide an important reference in establishing modern nuclear reactor safety at the atomic level.

We constructed a radioactive isotope (RI) separator named ERIS (electron-beam-driven RI separator for SCRIT) for the SCRIT (Self-Confinement RI Target) electron scattering facility at RIKEN RI Beam Factory (RIBF). In ERIS, production rate of fission products in the photofission of uranium is estimated to be 2.2 × 10 11 fissions/s with 30 g of uranium and a 1-kW electron beam. During the commissioning of ERIS, the mass resolution and overall efficiency, including ionization, extraction, and transmission, were found to be 1660 and 21%, respectively, using natural xenon gas. The preparation of uranium carbide (UC 2 ) RI production targets is described from which a 132 Sn beam was successfully separated in our first attempt at RI production.

Lanthanum carbide based targets were produced as benchmark tests before the production of radioactive uranium carbide targets. Carbides possessing excess carbon and porosity seem to be the best candidates as target for the production of exotic beams in the SPES-ISOL facility. In addition, the capability of tailoring properties such as grains size and pores size represents a step ahead to improve the ions release efficiency. In this work, multilayered graphene was used as source of carbon for the production of LaCx and the main physical properties of the produced targets were compared to standard LaCx produced using micrometric graphite. The main output of the work consisted in the reduced total porosity (28.8 vol% vs 47.8 vol%) and increased shrinkage (20.4 vol% vs 5.8 vol%) of the LaCx-Graphene samples compared to LaCx-Graphite ones. This result showed how graphene can be successfully employed as sintering aid for the sintering of carbides. Further studies are ongoing with UO2 as starting reagent for carburization within the project AUL-2013-16-176 “Study of the use of graphene as source of carbon for Uranium Carbide-Graphene nanocomposites production” now under conclusion at the European Commission DG Joint Research Centre – JRC.

International audience; $In\ situ$ 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 to 848 K and in an atmosphere of 10 to 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 to 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 $\mu$m/s.

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U3O8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significantmore » quantities of PuCl3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl3 by the precipitation of PuPO4.« less

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

International audience; In the context of radioactive ion beams, fission targets, often based on uranium compounds, have been used for more than 50 years at isotope separator on line facilities. The development of several projects of second generation facilities aiming at intensities two or three orders of magnitude higher than today puts an emphasis on the properties of the uranium fission targets. A study, driven by Institut de Physique Nucléaire d'Orsay (IPNO), has been started within the SPIRAL2 project to try and fully understand the behavior of these targets. In this paper, we have focused on five uranium carbide based targets. We present an off-line method to characterize their fission product release and the results are examined in conjunction with physical characteristics of each material such as the microstructure, the porosity and the chemical composition.

International audience; In this work, an experimental study of the radiance of liquid and solid uranium and plutonium carbides at wavelengths 550 nm ⩽ Î" ⩽ 920 nm is reported. A fast multi-channel spectro-pyrometer has been employed for the radiance measurements of samples heated up to and beyond their melting point by laser irradiation. The melting temperature of uranium monocarbide, soundly established at 2780 K, has been taken as a radiance reference. Based on it, a wavelength-dependence has been obtained for the high-temperature spectral emissivity of some uranium carbides (1 ⩽ C/U ⩽ 2). Similarly, the peritectic temperature of plutonium monocarbide (1900 K) has been used as a reference for plutonium monocarbide and sesquicarbide. The present spectral emissivities of solid uranium and plutonium carbides are close to 0.5 at 650 nm, in agreement with previous literature values. However, their high temperature behaviour, values in the liquid, and carbon-content and wavelength dependencies in the visible-near infrared range have been determined here for the first time. Liquid uranium carbide seems to interact with electromagnetic radiation in a more metallic way than does the solid, whereas a similar effect has not been observed for plutonium carbides. The current emissivity values have also been used to convert the measured radiance spectra into real temperature, and thus perform a thermal analysis of the laser heated samples. Some high-temperature phase boundaries in the systems Uâ€"C and Puâ€"C are shortly discussed on the basis of the current results.

Mixed Uranium-Plutonium Carbides are candidate fuels for 4th generation nuclear reactors. In this frame, a research program is underway at CEA (French alternative and atomic energies commission) to gather basic properties of such material to get a better understanding of in-pile behaviour. In the field of solid state physics, the measurement of transport properties and in particular diffusion coefficient requires the preparation of uranium carbide samples with very specific characteristics: very low oxygen content, mono-phased UC, high density (greater than 95% of the theoretical one), coarse microstructure, accurate samples geometry. In this paper the Process development is presented to manufacture such samples and the specific equipments which have been set up in glove-box. Characterizations of the first samples are also given.

The dissolution of uranium carbide pods as surrogate of UC fuel was performed by adding calcium metal in nitric acid, individually or in conjunction with oxide WWER's fuel. Extraction tests realized with the solution coming from the dissolution of the uranium carbide pods under PUREX conditions demonstrated that the plutonium remained in the aqueous phase. According to literature, it was found that it is connected with the formation of water-soluble organic compounds during the disslution. Extraction test performed on solutions coming from the dissolution a mix of uranium-carbide and oxide fuel (WWER) showed a complete extraction of uranium and plutonium and no plutonium retention of in the organic phase at the stripping step.

For the purpose of reprocessing of irradiated nuclear fuel from the water-cooled graphite-moderated pressure-tube reactor named AMB from decomissioned Russian “Atom Peaceful Big”, modernization of the process flow-sheet of the RT-1 plant is being carried out at PA Mayak with participation of FSUE KRI and VNIINM. A particular AMB SNF feature is extremely broad range of fuel compounds with the main ones being the uranium-molybdenum metal, uranium oxide and uranium carbide compositions usually dispersed in magnesium or calcium. Wide range of fuel compositions required to amend SNF dissolution, extraction processing, evaporation of high-level radioactive wastes and vitrification of high-level radioactive wastes. The above set of laboratory research was completed with dynamic tests using samples of AMB from the water-cooled graphite-moderated pressure-tube reactor. Tests have shown the possibility of processing the entire range of AMB SNF at the radiochemical plant RT-1 plant of the PA Mayak. Thus, the ability of the RT-1 plant to process different fuel compositions, including the long-term research reactor fuel have been proved experimentally.

A Phase Field model is developed combining the Orientation Field approach to modeling solidification with the Kim, Kim, Suzuki method of modeling binary alloys. These combined methods produce a model capable of simulating randomly oriented second phase dendrites with discrete control of the solid-liquid interface energy and thickness. The example system of carbon in a liquid uranium (U) melt is used as a test for the model. The formation of uranium carbide within a liquid U melt is simulated for isothermal conditions and compares well with experiments.

article i nfo Article history: Received 11 May 2010 Accepted 4 September 2010 Available online xxxx The detection of metallic silver on Chemical Vapour Deposited (CVD) grown silicon carbide and in Pebble Bed Modular Reactor (PBMR) supplied tri-structural isotropic (TRISO) coated particles (with 500 μm diameter zirconium oxide surrogate kernel) has been studied with femtosecond Laser Induced Breakdown Spectroscopy (femto-LIBS). The SiC layer of the TRISO coated particle is the main barrier to metallic and gaseous fission products of which 110m Ag is of particular interest for direct cycle high temperature reactors. This work is a feasibility study for diagnosing and profiling silver transport through the silicon carbide layer of fuel particles for a high temperature gas reactor in out-of-reactor experimentation. The zirconium oxide is a surrogate for the enriched uranium oxide fuel. The conclusion reached in this study was that femto-LIBS can achieve good surface spatial resolution and good depth resolution for studies of silver in experimental coated particles. The LIBS technique also offers a good alternative for a remote analytical technique.

Uranium carbide dispersed in graphite was produced under vacuum by means of carbothermic reduction of different uranium oxides (UO2, U3O8 and UO3), using graphite as the source of carbon. The thermal process was monitored by mass spectrometry and the gas evolution confirmed the reduction of the U3O8 and UO3 oxides to UO2 before the carbothermic reaction, that started to occur at T > 1000 °C. XRD analysis confirmed the formation of α-UC2 and of a minor amount of UC. The morphology of the produced uranium carbide was not affected by the oxides employed as the source of uranium.

The US Department of Energy plans to use the internal gelation process to make tristructural isotropic (TRISO)-coated transuranic (TRU) fuel particles. The focus of this work is to develop TRU fuel kernels with high crush strengths, good ellipticity, and adequately dispersed silicon carbide (SiC). The submicron SiC particles in the TRU kernels are to serve as getters for excess oxygen and to potentially sequester palladium, rhodium, and ruthenium, which could damage the coatings during irradiation. Zirconium oxide microspheres stabilized with yttrium were used as surrogates because zirconium and TRU microspheres from the internal gelation process are amorphous and encounter similar processing problems. The hardness of SiC required modifications to the experimental system that was used to make uranium carbide kernels. Suitable processing conditions and equipment changes were identified so that the SiC could be homogeneously dispersed in gel spheres for subsequent calcination into strong spherical kernels.

In the context of the material research aimed at supporting the development of nuclear plants of the fourth Generation, renewed interest has recently arisen in carbide fuels. A profound understanding of the behaviour of nuclear materials in extreme conditions is of prime importance for the analysis of the operation limits of nuclear fuels, and prediction of possible nuclear reactor accidents. In this context, the main goal of the present paper is to demonstrate the feasibility of laser induced melting experiments on stoichiometric uranium carbides; UC, UC1.5 and UC2. Measurements were performed, at temperatures around 3000 K, under a few bars of inert gas in order to minimise vaporisation and oxidation effects, which may occur at these temperatures. Moreover, a recently developed investigation method has been employed, based on in situ analysis of the sample surface reflectivity evolution during melting. Current results, 2781 K for the melting point of UC, 2665 K for the solidus and 2681 K for the liquidus of U2C3, 2754 K for the solidus and 2770 K for the liquidus of UC2, are in fair agreement with early publications where the melting behaviour of uranium carbides was investigated by traditional furnace melting methods. Further information has been obtained in the current research about the non-congruent (solidus–liquidus) melting of certain carbides, which suggest that a solidus–liquidus scheme is followed by higher ratio carbides, possibly even for UC2.

The knowledge of thermal radiative and thermodynamic properties of uranium and plutonium carbides under extreme conditions is essential for designing a new metallic fuel materials for next generation of a nuclear reactor. The present work is devoted to the study of the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides at their melting/freezing temperatures. The Stefan-Boltzmann law, total energy density, number density of photons, Helmholtz free energy density, internal energy density, enthalpy density, entropy density, heat capacity at constant volume, pressure, and normal total emissivity are calculated using experimental data for the frequency dependence of the normal spectral emissivity of liquid and solid uranium and plutonium carbides in the visible-near infrared range. It is shown that the thermal radiative and thermodynamic functions of uranium carbide have a slight difference during liquid-to-solid transition. Unlike UC, such a difference between these functions have not been established for plutonium carbide. The calculated values for the normal total emissivity of uranium and plutonium carbides at their melting temperatures is in good agreement with experimental data. The obtained results allow to calculate the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides for any size of samples. Based on the model of Hagen-Rubens and the Wiedemann-Franz law, a new method to determine the thermal conductivity of metals and carbides at the melting points is proposed., Comment: 17 pages, 2 Tables

Scoping calculations have been performed for a very high temperature (1000 °C) helium-cooled fast reactor involving two distinct options: (1) using graphite foam into which UC (12% enrichment) is embedded into a matrix comprising UC and graphite foam molded into hexagonal building blocks, shown in Fig. 1, and encapsulated with a SiC shell covering all surfaces, and (2) using UC only (also 12% enrichment) molded into the same shape and size as the foam–UC matrix in option 1. Both options use the same basic hexagonal fuel matrix blocks to form the core and reflector. The reflector contains natural uranium only. Both options use 50 μm SiC as a containment shell for fission product retention within each hexagonal block. The calculations show that the option using foam (option 1) would produce a reactor that can operate continuously for at least 25 years without ever adding or removing any fuel from the reactor. The calculations show further that the UC only option (option 2) can operate continually for 50 years without ever adding or removing fuel from the reactor. Doppler and loss of coolant reactivity coefficients were calculated. The Doppler coefficient is negative and much larger than the loss of coolant coefficient, which was very small and positive. Additional progress on and development of the two concepts are continuing.

Numerical simulation is used to study the possibility of long-term chain nuclear reactions deep inside the Earth over 4 billion years. The active layer (the natural nuclear reactor operating on fast neutrons in lakes) can form when uranium oxides or carbides precipitate from a liquid layer onto the Earth’s solid interior core. A mechanism of uranium concentration in the Earth’s core is studied, experiments are performed, and it is shown that a nuclear chain reaction with breeding of fissioning nuclides could have occurred in such a layer. The basic neutron-physical characteristics of such a natural nuclear reactor are calculated. It most likely operates in a pulsed mode. The critical condition for the duration of the reaction is the power level. It is found that for certain optimal power this process in the Earth’s core can last for more than 4 billion years up to the present time.

A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO2), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO2/UO2), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials ...

Radioactive Ion Beams (RIBs) at the Holifield Radioactive Ion Beam Facility (HRIBF) are produced using the isotope separation on-line technique and are subsequently accelerated up to a few MeV per nucleon for use in nuclear physics experiments. The first RIB experiments at the HRIBF were completed at the end of 1998 using 17F beams. Since then other proton-rich ion beams have been developed and a large number of neutron-rich ion beams are now available. The neutron-rich radioactive nuclei are produced via proton-induced fission of uranium in a low-density matrix of uranium carbide. Recently developed RIBs include 25Al from a silicon carbide target and isobarically pure beams of neutron-rich Ge, Sn, Br and I isotopes from a uranium carbide target.

It is shown that during annealing of uranium samples at 550–950°C in a 10–3 Pa vacuum carbon impurity diffuses from the interior volume to the surface of the samples. The diffusion flux of carbon directed toward the surface arises as a result of the formation of uranium oxymonocarbide, which contains carbon impurity, on the surface in the presence of low oxygen partial pressure. The carbon diffusion rate in α-uranium grains depends on the crystallographic orientation of the grains, i.e., carbon diffusion in α-uranium is anisotropic.