Uranium Mononitride Research Articles

Radiation induced athermal diffusivity in uranium mononitride

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuel considered for light water reactors and advanced reactor designs. Properties like self- and fission gas diffusivity need to be better understood, given that they influence key fuel performance phenomena such as fission gas swelling and release. In particular, the radiation induced athermal (D3) diffusivity remains challenging to accurately predict and has only been sparsely characterized in UN, despite its importance as it likely governs diffusion at the low temperatures this high-thermal-conductivity fuel form may operate. Molecular Dynamics simulations are used to estimate the mean square displacement induced by a primary knock-on atom (PKA) with a given kinetic energy. These results are combined with the PKA energy distributions obtained from binary collision approximation calculations to obtain the displacement due to a particular fission fragment. Finally, this is combined with experimental fission fragment yields to determine the displacement due to an average fission event and, thus, express the athermal diffusivity as a function of the fission rate density. These results are in excellent agreement with available experimental data. A particular importance is given to the understanding and the quantification of the variability of these results.

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/ 10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

Assessment of uranium nitride interatomic potentials

Uranium mononitride (UN) is a promising nuclear fuel due to its high fissile density, high thermal conductivity, and suitability for reprocessing. In this study, two uranium nitride interatomic potentials are assessed: Tseplyaev and Starikov's angular-dependent potential and Kocevski et al.'s embedded atom model potential. Predictions of the thermophysical and elastic properties of UN, ▪, and α- and β-▪ computed using both potentials are assessed and compared to available experimental data. The Tseplyaev potential performs better with the energetic aspects of UN, e.g., specific heat capacity and point defect formation energies, whereas the Kocevski potential performs better with the structural aspects of UN, e.g., thermal expansion as well as with the elastic properties. The reasons why the Kocevski potential underestimates the UN specific heat are explained by examining the UN phonon properties modeled using both potentials. The Kocevski potential shows better identification of the mechanical stability ranges of UN, ▪, and α- and β-▪, reasonably predicting the melting point of UN and predicting stable structures for ▪ and α- and β-▪. On the other hand, the Tseplyaev potential predicts a premature phase change of both UN and ▪ and cannot stabilize α- nor β-▪. However, the Kocevski potential cannot predict a stable α-U phase and is thus not suitable for the calculation of formation energies for non-stoichiometric point defects.

Preparation of UN microspheres by external gelation

Uranium mono-nitride (UN) microspheres are expected to play an important role in various miniaturization module reactors. The combination of UN core and tri-structural-isotropic (TRISO) coatings will bring improved efficiency and safety. Carbon-containing gel microspheres were prepared by external gelation, followed by carbothermal reduction and nitridation to synthesize UN. The effects of C/U molar ratios and heat-treatment schemes on the composition and microstructures of UN microspheres were studied. The results indicated that the external gelation could be used to prepare UN microspheres, with the highest UN content of 93 wt%.

Resilience of uranium mononitride/zirconium carbide composites and uranium-zirconium carbonitride in hot hydrogen for nuclear thermal propulsion

Nuclear thermal rockets require fuels capable of withstanding flowing hydrogen propellant up to 3200 K. Presently, there has not been a fuel type that reliably operates at these conditions. A promising candidate anticipated to endure this demanding environment is a ceramic-ceramic composite comprising of uranium mononitride and zirconium carbide. This investigation assesses the behavior and resilience of variations of this composite and its resultant homogenized form (uranium-zirconium carbonitride) under two hot hydrogen conditions (2273 K and 3000 K). The findings revealed that composites that homogenize into UZrCN exhibit superior structural integrity in hydrogen compared to heterogeneous counterparts. Consequently, this study underscores the potential of homogenized uranium-zirconium carbonitride for enhanced performance in nuclear thermal propulsion applications.

Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix

Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a metallic matrix (cermet) has been proposed. To assess the viability of this concept, hot hydrogen testing of tungsten-coated UN kernels embedded in a Mo-30 wt% W (Mo30W) alloy matrix has been performed at temperatures from 1800°C to 2300°C. Both the isolated kernels and kernels consolidated by spark plasma sintering in the Mo30W matrix were tested. In addition to direct observations and mass loss measurements, the samples were analyzed by X-ray diffraction (XRD) and scanning electron microscopy (SEM)/energy dispersive X-ray spectroscopy (EDS) after each run. The decomposition of UN started at 1800°C despite the coating and matrix, and increased at 2000°C. Uranium seeped through the tungsten grain boundaries of the coating at all temperatures. The consolidated sample expanded irregularly at 2000°C through the formation of voids, and SEM/EDS analysis showed uranium-containing veins in the matrix consisting of U2Mo according to the XRD data. The observed pore generation at 2000°C was explained by the formation of water vapor from residual oxides and diffused hydrogen. At 2200°C and above, both the kernels and the consolidated samples melted through the formation of uranium or low–melting point uranium-molybdenum alloys.

First-principles analysis of intergranular fracture in UN by Σ5(210) grain boundary segregated Xe and vacancy

Segregation of irradiation-generated xenon (Xe) atoms and vacancies to grain boundary (GB) leads to instability and brittleness of uranium mononitride (UN). However, theoretical knowledge about the energetics of its defective interface is lacking, whereas there has been considerable attention on unirradiated UN in experiments. Using first-principles methods, herein we compute the energies of segregation and fracture for typical fission defects. We find that, compared with vacancy, Xe tends more to segregate to Σ5(210) GB, enhancing sensitivity to intergranular fracture. Our predictions clarify the fundamental embrittlement mechanism of irradiated UN GB, which has been discussed so far only on pristine one.

Simulation of uranium mononitride spent fuel: A thermodynamic approach

Uranium mononitride (UN) is a potential versatile fuel for use in both thermal and fast spectrum reactors. Knowledge of the thermodynamic properties of actinides and fission products in spent UN fuel is required to understand their properties such as phase stability and retention during long term storage and disposal. The present study reviews thermodynamic data and calculates the free energies of formation (ΔfGm) of the nitrides and other components formed in the spent fuel to predict the actinide and fission product behaviour.An End of Life (EoL) spent fuel inventory was calculated for a high burnup UN fuel (60 MWd kg−1) using the FISPIN fuel inventory code. The spent fuel consisted predominantly of a solid solution of nitrides (U, An, Ln, Y, Zr, Nb)N forming a single homogeneous and stable phase as expected from the ΔfGm variation with T of its components or of the fuel as calculated for an ideal solid solution. Since δΔfGm/δT > 0 for the reduction: MN ⇔ M + ½ N2, dissociation is consequently more likely at high temperature.Other fission products are expected to be divalent (Ba, Sr), monovalent (Cs, Rb) or non-valent (Tc, Ru, Rh, Pd) as well as noble gases (Xe, Kr), and halides (I, Br) which may form nano-precipitates (e.g. with metal ions). The behaviour of Mo is more complex. At low fuel temperature (<1100 K) it may form nitride precipitates while at higher temperature (>1100 K) MoN and MoN0.5 decompose to Mo metal. The stoichiometry of the spent fuel is related to the burn-up and the temperature of the fuel during operation. It is also shown to be dependent on the molybdenum species generated in pile (metal or nitride precipitate type).Thermodynamic calculations for potential Pellet Clad Interaction (PCI) showed that with Zr alloys interactions are expected while with stainless steel clad no reaction between the steel components (Fe, Ni, and Cr) and UN is expected. Finally, the free energy evaluation for UN hydrolysis shows that UN reaction with water is spontaneous.

Phase stability of uranium monocarbide under laser irradiation in argon and methane environments

Laser irradiation is becoming a technique of choice to study materials behavior under extreme conditions due to its ability of achieving very high temperatures within minute times. Recent studies have indicated that laser irradiation can be used to synthesize inorganic compounds from various organic compounds that usually decompose or oxidize if conventional heat treatment techniques are used. It was also shown that metallic uranium can be synthesized if uranium mononitride is laser irradiated under inert environments. This current study focused on the high temperature (in the order of 2500 K) behavior of laser-irradiated uranium monocarbide (UC) using a 1070 nm laser with a nominal spot size of 37 µm and 50–250 W power and density of 139 – 694 Wcm−2 under argon (Ar(g)) and methane (CH4(g)) environments. Laser irradiation of UC at 50 W did not show significant UC decomposition under Ar(g) where no change in the concentration of UO2 or formation of oxygen-dissolved carbide phases were observed. However, the presence of secondary oxygen-dissolved phases such as UC1-xOx, assuming stoichiometric compositions, and UC2-yOy in UC samples laser irradiated under a similar environment at 50–250 W suggested a slight evolution of carbon from UC in the presence of impurity oxygen. Up to 70 wt.% dicarbide (α-UC2) phase formed when UC was laser irradiated up to 50 W under CH4(g). These observations suggest high thermal stability of UC at macroscale under inert atmosphere in the absence of oxygen impurities and a high reactivity of UC with carbon-rich environment up to 50 W laser irradiation. Crystallographic properties such as lattice strain and crystallite sizes of these different phases showed complicated variation without significant correlation to the laser irradiation conditions, and these observations are also discussed here.

The effects of Zr on incorporation, diffusion, and clustering behaviors of He in uranium mononitride: A first-principles study

The effects of Zr on incorporation, diffusion, and clustering behaviors of He in uranium mononitride: A first-principles study

Simulations of self- and Xe diffusivity in uranium mononitride including chemistry and irradiation effects

A combination of density functional theory and empirical potential atomic scale simulations have been used to determine a model for defect stability and mobility in uranium mononitride (UN), as a function of temperature (T) and N2 partial pressure (pN2). Using the model, predictions of hypo-stoichiometry under U-rich conditions compare favorably to CALPHAD calculations using the TAF-ID database. Furthermore, our predictions of U and N self-diffusivity are in good agreement with experiments carried out as a function of T at specific partial pressures under thermal equilibrium. The validated atomic scale data have then been implemented within a cluster dynamics method to simulate irradiation-enhanced defect concentrations. All defects and clusters studied have significantly enhanced concentrations, with respect to thermal equilibrium, as T is lowered. The irradiation-enhanced Xe diffusivity is compared to post-irradiation annealing and in-pile experiments. The contributions of various defects and clusters to non-stoichiometry, self-diffusivity, and Xe diffusivity are discussed.

Neutronic and thermal hydraulic analysis of a PWR fuel assembly using uranium mononitride

The nitride fuel is currently considered as a material to replacement of uranium dioxide in Light Water Reactors (LWR), mainly due to its higher thermal conductivity when compared with the oxide fuels and similar melting point. This behavior may contribute to larger heat conduction in the fuel assemblies and contributes to increase the safety margins to fuel melting during operation. In this way, the present work aims to contribute to studies on the use of uranium mononitride (UN) in LWRs. Neutronic and thermal hydraulic analysis were performed for a typical assembly of Pressurized Water Reactor (PWR) using uranium dioxide (UO2) and uranium mononitride (UN) fuels. The SCALE 6.0 (Standardized Computer Analyses for Licensing Evaluation – version 6.0) and STHIRP (Thermal Hydraulic Simulation of Research Reactors) codes were used to calculate the neutronic and thermal hydraulic parameters, respectively. The goal is to verify the possible use of UN in a PWR. Comparing the evaluated fuels, the largest thermal conductivity of UN contributes to largest heat conduction in the pellet zone and to lowest fuel temperature. However, the UN presents highest production of transuranic elements and considerable 14C production into UN during the burnup. Results of neutronic and thermal hydraulic parameters behavior for both fuels are presented in details in this work as well as carefully analyzed.

The incorporation of xenon at point defects and bubbles in uranium mononitride

Uranium mononitride (UN) has been proposed as an accident tolerant fuel for nuclear fission reactors and offers enhanced performance during accident scenarios relative to the current fuel, uranium dioxide. However, its performance in reactor is significantly less well understood than for the oxide. Therefore, this work explores incorporation of Xe into UN using density functional theory to understand the early stages of fission gas evolution. These results are used to derive a new potential for Xe in UN, which is then employed to simulate the growth of xenon bubbles in spherical voids of various sizes at 300 K and 1200 K. At sufficiently high gas densities, the xenon was found to mainly crystallise in an fcc arrangement. Loop punching was observed at 10.2 GPa and above for larger bubbles of 4.8 nm radius, significantly so for higher temperatures. This work suggests that no Xe undergoes thermal resolution at temperatures up to 1200 K and that the UN lattice prefers to undergo deformation instead.

High resolution electronic spectroscopy of uranium mononitride, UN.

The isoelectronic molecules UN and UO+ are known to have Ω = 3.5 and Ω = 4.5 ground states, respectively (where Ω is the unsigned projection of the electronic angular momentum along the internuclear axis). A ligand field theory model has been proposed to account for the difference [Matthew and Morse, J. Chem. Phys. 138, 184303 (2013)]. The ground state of UO+ arises from the U3+(5f3(4I4.5))O2- configuration. Owing to the higher nominal charge of the N3- ligand, the U3+ ion in UN is stabilized by promoting one of the 5f electrons to the more polarizable 7s orbital, reducing the repulsive interaction with the ligand and rendering U3+(5f27s(4H3.5))N3- the lowest energy configuration. In the present work, we have advanced the characterization of the UN ground state through studies of two electronic transitions, [18.35]4.5-X(1)3.5 and [18.63]4.5-X(1)3.5, using sub-Doppler laser excitation techniques with fluorescence detection. Spectra were recorded under field-free conditions and in the presence of static electric or magnetic fields. The ground state electric dipole moment [μ = 4.30(2) D] and magnetic ge-factor [2.160(9)] were determined from these data. These values were both consistent with the 5f27s configurational assignment. Dispersed fluorescence measurements were used to determine vibrational constants for the ground and first electronically excited states. Electric dipole moments and magnetic ge-factors are also reported for the higher-energy electronically excited states.

Diffusion study of uranium mononitride/zirconium carbide composite for space nuclear propulsion

The next generation of space exploration will require extensive developments in rocket technology. Space nuclear propulsion is of interest due to its high fuel density and power; however, it also has high temperature and stability demands. This study examines a uranium mononitride (UN) and zirconium carbide (ZrC) ceramic-ceramic particle composite as a potential fuel for these missions. Based on the constituent properties, this fuel composite is expected to be thermally efficient and resistant to the hot hydrogen propellant. One of the main concerns with this composite is the unknown diffusion behavior between UN and ZrC over time. Diffusion couple experiments and isothermal furnace tests of composites were two experimental techniques used to determine how these constituents will behave when left in contact at high temperatures. UN and ZrC were found to have limited but observable diffusion at the phase boundary. The resulting phase is a UZr(CN) quaternary phase. UZr(CN) has been examined for other high temperature and gas nuclear reactors with preliminary success; however, there lacks sufficient data to fully understand this phase. Based on the limited information available, the resultant quaternary phase could be deemed acceptable and even provide better adhesion for the fuel particles (UN) to the matrix (ZrC).

Finite temperature properties of uranium mononitride

Uranium mononitride (UN) is a promising nuclear fuel that combines the advantageous properties of readily used UO2 and uranium alloys, such as high melting temperature and high uranium density, and thermal conductivity, respectively. A better understanding of UN behavior at operating temperatures can be obtained from finite temperature data, such as elastic properties. To get this information, ab initio molecular dynamics (AIMD) simulations were performed at five different temperatures using constant volume (NVT) and constant pressure (NPT) ensembles. Initially, the performance of PBE functional in reproducing experimental crystallographic properties and magnetic ordering is assessed. The finite temperature phonon dispersions are calculated using NVT simulation results, which show a softening of the phonon modes with increasing temperature. The NPT results are used to obtain the thermal expansion of UN and finite temperature electronic properties. The calculated thermal expansion is compared with our measurements using neutron diffraction. Additionally, the temperature dependent elastic properties of UN are evaluated using the strain-stress method in AIMD simulations, indicating that UN becomes softer and more compressible with increasing temperature. Also, the calculated Young’s modulus slope is in very good agreement with the experiment. The finite temperature heat capacity and electronic thermal conductivity are calculated from AIMD simulations, which are in better agreement with the experiment than the heat capacity and thermal conductivity calculated using the structures relaxed at 0 K. Lastly, the thermal diffusivity from AIMD has opposite temperature dependence compared to experimental results, which we argued comes from the underestimated electronic thermal conductivity.

TREAT testing of additively manufactured SiC canisters loaded with high density TRISO fuel for the Transformational Challenge Reactor project

Specimens composed of a novel nuclear fuel architecture underwent testing in the Transient Reactor Test Facility (TREAT) to investigate their behavior in thermomechanical states only possible with this type of testing. Uranium mononitride fuel kernels, with coatings typical of Tristructural Isotropic (TRISO) systems, were loaded into additively manufacturing silicon carbide (SiC) canisters and treated by chemical vapor infiltration so that the TRISO particles were held in a fully ceramic SiC matrix. A metallic heat sink capsule was used in TREAT to test fresh fuel specimens under progressively higher energy depositions. Thermomechanical modeling and post transient examinations, including a unique application of x-ray computed tomography, showed that more energetic transients led to stronger temperature gradients, higher stresses, and increased level of fracturing in the specimens. While some specimens were subjected to challenging conditions, the level of fracturing observed did not show a measurable dimensional change or loss of specimen geometry. These results indicate that this fuel architecture has resilience to catastrophic failure under power pulse transients. Future experiments are recommended for specimens which have been previously irradiated in order to investigate the effects of fuel burnup and neutron fluence on transient fuel performance.

Thermal expansion and steam oxidation of uranium mononitride analysed via in situ neutron diffraction

In situ neutron powder diffraction experiments are applied to physical, kinetic, and microstructural characterization of uranium mononitride as a promising light water reactor fuel material. The temperature-variable coefficient of thermal expansion and isotropic Debye Waller factors are obtained by sequential Rietveld refinement over 499–1873 K. Oxidation of a UN pellet (95.2% density) under flow of 11 mg/min D2O is observed to initiate above 623 K and the rate increases by a factor of approximately 10 from 673 to 773 K, with activation energy 50.6 ± 1.3 kJ/mol; uranium oxide is the only solid corrosion product.

High local oxygen coverage causes initial oxidation of UN(001) surface

Uranium mononitride (UN) is a promising nuclear fuel, however, during its storage and operation, the surface oxidation or the initial corrosion of UN always takes place. The limited understanding of atomistic corrosion mechanism hinders the improvement of UN corrosion resistance. Here, the interactions between adsorbates (O2 and H2O molecules and O adatoms) and the pristine UN(001) surface, the UN(001) surface with N vacancies (Nvac_UN(001)), and the UN(001) surface with N vacancies occupied by O atoms (O2N_UN(001)) are investigated by using density functional theory calculations. O2 molecule is found to spontaneously dissociate on these surfaces, while the H2O molecule is weakly adsorbed on the top site of U atom. Due to the attractive OO lateral interactions, high local O coverage (e.g., 4O coordinated to a U atom) is easily resulted, where significant relaxation of surface U atoms is observed. The maximum vertical displacement of U atoms is more than 1 Å at medium O coverage (e.g., 3O coordinated to a U atom). N vacancy and O-doping on the UN(001) surface do not change the vertical displacement of U atoms, but alter the adsorption energies of O atoms. The competition between surface U-O and subsurface U-N interactions leads to the formation of uranium cavities in the surface, preluding the dissolution as oxides of U atoms away from the surface. Our results provide new insights for the understanding of initial oxidative corrosion of UN surface.

A Rapid Sintering Method for Cerium Nitride Pellet: A Uranium Mononitride Surrogate

Uranium mononitride (UN) is a candidate fuel material for light water reactors with higher uranium (U) loading and thermal conductivity than uranium dioxide (UO2). However, the sintering of UN pellets is challenging as the UN powder particles oxidize rapidly at high temperatures unless the oxygen concentration is extremely low. Oxidation during sintering either reduces the relative density of the sintered UN pellet or disintegrates the sintered UN pellet to powder. To address this problem, the present work developed a rapid sintering method for producing highly densified UN surrogate pellets with minimal oxidation. Cerium nitride (CeN) is used as a surrogate for UN to reduce radiation hazards. With the custom-developed fast-heating system, the sintering process was completed within 150 s. The sintering atmosphere was flowing nitrogen (N2). The sintered CeN pellet density was 95% of the theoretical density (TD) or higher. The microstructure was uniform with a 10–25 µm grain size as demonstrated by scanning electron microscopy (SEM) and contained trivial levels of oxides as demonstrated by X-ray diffraction (XRD). The resultant pellets indicate that the rapid sintering method is a promising method to make UN fuel pellets with equivalent or higher density to pellets made by conventional sintering methods, while also being more efficient in time and costs.