Uranium Nitride Research Articles

Oxidation studies of UM3 (M = Ru, Rh, Pd) intermetallides

Mixed uranium and plutonium nitride (U,Pu)N is a promising candidate fuel for fast reactors. Fission of U and Pu during reactor operation is accompanied by the formation of chemically inert compounds of uranium and noble metals - intermetallides (IMCs) with the general formula UM3 (M = Ru, Rh, Pd). One of possible steps in reprocessing of spent nitride nuclear fuel is voloxidation - oxidation in an oxygen-containing environment aimed at removal of gaseous components of SNF. In contrast to well-studied oxidation of uranium nitride in air at high temperatures, the oxidation of intermetallic compounds of uranium and noble metals has not been studied in sufficient detail. In the present work, it is shown that UM3 (M = Ru, Rh, Pd) IMCs are completely oxidized in the temperature range 800–1200 °C. UPd3 and URu3 oxidation leads to the formation of U3O8, PdO and RuO2. Upon oxidation of URh3, the formation of an intermediate phase Rh2UO6 is observed; upon further heating, it decomposes into U3O8 and metallic Rh. It is shown that peculiarities of oxidation of the studied IMCs are determined by reactivities of relevant noble metals with respect to oxygen. This study demonstrates that voloxidation of IMCs from nitride SNF should be performed at temperatures above 1200 °C.

Coated ZrN sphere-UO2 composites as surrogates for UN-UO2 accident tolerant fuels

Uranium nitride (UN) spheres embedded in uranium dioxide (UO2) matrix is considered an innovative accident tolerant fuel (ATF). However, the interaction between UN and UO2 restricts the applicability of such composite in light water reactors. A possibility to limit this interaction is to separate the two materials with a diffusion barrier that has a high melting point, high thermal conductivity, and reasonably low neutron cross-section. Recent density functional theory calculations and experimental results on interface interactions in UN-X-UO2 systems (X = V, Nb, Ta, Cr, Mo, W) concluded that Mo and W are promising coating candidates. In this work, we develop and study different methods of coating ZrN spheres, used as a surrogate material for UN spheres: first, using Mo or W nanopowders (wet and binder); and second, using chemical vapour deposition (CVD) of W. ZrN-UO2 composites containing 15 wt% of coated ZrN spheres were consolidated by spark plasma sintering (1773 K, 80 MPa) and characterised by SEM/FIB-EDS and EBSD. The results show dense Mo and W layers without interaction with UO2. Wet and binder Mo methods provided coating layers of about 20 µm and 65 µm, respectively, while the binder and CVD of W methods layers of about 12 µm and 3 µm, respectively.

Preliminary conceptual design of fast neutron spectrum nuclear thermal rocket cores using monolithic uranium nitride fuel

This paper presents a few nuclear thermal rocket (NTR) reactor core preliminary conceptual designs based on the use of monolithic uranium nitride (UN) fuel plates or pins clad with tungsten (W), alloyed or unalloyed as needed. High-assay low-enriched uranium (HALEU) as well as high-enriched uranium (HEU) are considered. Nominal core thermal powers are between 300 and 1100 MW corresponding to thrusts between 66 kN and 240 kN. The estimated thrust-to-reactor-weight ratios (T/WRx) achievable with the HEU plate and pin configurations are between 3.8 and 10.9. The T/WRx achievable with the HALEU plate and pin configurations are lower, 1.9–4.0, than those achievable with HEU. It is noteworthy that even a relatively modest increase in fuel enrichment above the HALEU limit enables a significant increase in T/WRx. It must be emphasized that these results do not yet account for uncertainties and that the appropriateness of the assumed fuel thermal design limits for normal operation (i.e., peak centerline temperature of 3100 K) will need to be confirmed. Finally, just like for the other UN-based fuels currently considered for NTR applications (i.e., CERMET and CERCER), the proposed UN fuel systems need significant testing in prototypic conditions to confirm the appropriate performance for safe NTR applications. However, the monolithic fuel systems may have, overall, a higher technology readiness level (TRL), and, consequently, shorter fuel development and validation time may result.

Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of UN fuel reached when used heterogeneous core configuration case four with percentage of U-235 in F1=5.5 %, F2=7 % and F3=8.5 % also with the addition of Np-237 0.2 % and fuel fraction 56 %. It has a maximum excess reactivity value 12.56 % %∆k/k. And then, the optimum configuration of ThN fuel reached when used heterogeneous core configuration case three with percentage of U-233 in F1=2 %, F2=4 % and F3=6 % with the addition of Pa-231 0.5 % and fuel fraction 53 %. It has a maximum excess reactivity value 7.67 % %∆k/k. The comparison of optimum design of UN and ThN fuel shows that the ThN fuel has the k-eff value closer to critical than UN fuel. Therefore, in this study, ThN fuel is more suitable for use in PWR reactors because it has a small excess value and can operate for 10 years without refueling

Uranium: The Nuclear Fuel Cycle and Beyond.

This review summarizes the recent developments regarding the use of uranium as nuclear fuel, including recycling and health aspects, elucidated from a chemical point of view, i.e., emphasizing the rich uranium coordination chemistry, which has also raised interest in using uranium compounds in synthesis and catalysis. A number of novel uranium coordination features are addressed, such the emerging number of U(II) complexes and uranium nitride complexes as a promising class of materials for more efficient and safer nuclear fuels. The current discussion about uranium triple bonds is addressed by quantum chemical investigations using local vibrational mode force constants as quantitative bond strength descriptors based on vibrational spectroscopy. The local mode analysis of selected uranium nitrides, N≡U≡N, U≡N, N≡U=NH and N≡U=O, could confirm and quantify, for the first time, that these molecules exhibit a UN triple bond as hypothesized in the literature. We hope that this review will inspire the community interested in uranium chemistry and will serve as an incubator for fruitful collaborations between theory and experimentation in exploring the wealth of uranium chemistry.

High volume packing fraction TRISO-based fuel in light water reactors

For a decade, fully ceramic microencapsulated (FCM) fuel, containing tri-structural isotropic (TRISO) fuel particles in a silicon carbide (SiC) matrix, has been investigated as an accident-tolerant fuel for light water reactors (LWRs). Other examples exist of TRISO-based concepts for LWR fuels with different matrix materials. Previous studies assumed TRISO particle volume packing of approximately 0.44 in SiC (or another) matrix, the highest realistic packing fractions possible with conventional manufacturing. Recent advances in advanced manufacturing have yielded the development and demonstration of a fuel form that consists of conventionally manufactured TRISO particles in a 3D-printed SiC matrix with significantly higher possible TRISO packing fractions (0.5–0.7). This increased uranium loading enhances the viability of using TRISO-based particle fuel forms in LWRs. The viability of high-packing-fraction TRISO-based particle fuel forms in LWRs is assessed from the perspective of fuel cycle length, achievable fuel burnup, reactivity coefficients, and fuel cycle performance. Higher-packing-fraction TRISO-based fuel enables either longer cycle lengths (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant enrichment or decreased enrichments (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant cycle length. Studies of different fuel kernel types (uranium nitride, uranium oxycarbide, and uranium carbide) yield similar results, although the cycle length of uranium oxycarbide is shorter than for uranium nitride or uranium carbide (due to the lower density of uranium oxide). This work also characterized the production of 14C resulting from neutron absorption in 14N during operation for uranium mononitride fuel kernels; the ratio of 14C/N was 1–2 at. % at discharge. For the fuel cycle evaluation, the activity of spent nuclear fuel and high-level waste at 100 and 100,000 years was lower for high-packing-fraction fuels than for conventional LWR fuel. Environmental impact metrics were similar overall, but higher on the front end of the fuel cycle and lower on the back end of the fuel cycle. Reactivity coefficients of higher-packing-fraction TRISO-based fuel were reasonable compared with those of conventional fuels.

Uranium-nitride chemistry: uranium-uranium electronic communication mediated by nitride bridges.

Treatment of [UIV(N3)(TrenTIPS)] (1, TrenTIPS = {N(CH2CH2NSiPri3)3}3−) with excess Li resulted in the isolation of [{UIV(μ-NLi2)(TrenTIPS)}2] (2), which exhibits a diuranium(iv) ‘diamond-core’ dinitride motif. Over-reduction of 1 produces [UIII(TrenTIPS)] (3), and together with known [{UV(μ-NLi)(TrenTIPS)}2] (4) an overall reduction sequence 1 → 4 → 2 → 3 is proposed. Attempts to produce an odd-electron nitride from 2 resulted in the formation of [{UIV(TrenTIPS)}2(μ-NH)(μ-NLi2)Li] (5). Use of heavier alkali metals did not result in the formation of analogues of 2, emphasising the role of the high charge-to-radius-ratio of lithium stabilising the charge build up at the nitride. Variable-temperature magnetic data for 2 and 5 reveal large low-temperature magnetic moments, suggesting doubly degenerate ground states, where the effective symmetry of the strong crystal field of the nitride dominates over the spin–orbit coupled nature of the ground multiplet of uranium(iv). Spin Hamiltonian modelling of the magnetic data for 2 and 5 suggest U⋯U anti-ferromagnetic coupling of −4.1 and −3.4 cm−1, respectively. The nature of the U⋯U electronic communication was probed computationally, revealing a borderline case where the prospect of direct uranium–uranium bonding was raised, but in-depth computational analysis reveals that if any uranium–uranium bonding is present it is weak, and instead the nitride centres dominate the mediation of U⋯U electronic communication. This highlights the importance of obtaining high-level ab initio insight when probing potential actinide–actinide electronic communication and bonding in weakly coupled systems. The computational analysis highlights analogies between the ‘diamond-core’ dinitride of 2 and matrix-isolated binary U2N2.

Heterometallic uranium/molybdenum nitride synthesis via partial N-atom transfer.

The reaction of a terminal Mo(II) nitride with a U(III) complex yields a heterodimetallic U-Mo nitride which is the first example of a transition metal-capped uranium nitride. The nitride is triply bonded to U(V) and singly bonded to Mo(0) and supports a U-Mo interaction. This compound shows reactivity toward CO oxidation.

Calphad Modeling of Uranium Nitride (Un) Fabrication Routes Enabled by First-Principles Calculations

Calphad Modeling of Uranium Nitride (Un) Fabrication Routes Enabled by First-Principles Calculations

Cr-incorporated uranium nitride composite fuels with enhanced mechanical performance and oxidation resistance

Cr-incorporated UN (Cr-UN) composite fuels with different Cr amounts up to 10 wt% are fabricated by spark plasma sintering, and their microstructure and phase heterogeneity are analyzed. Highly densified microstructure and homogeneous Cr distribution are identified for the Cr-incorporated UN pellets, demonstrating a liquid sintering characteristic with a Cr enriched phase on the UN grain boundaries. A ternary phase U2CrN3 forms in the Cr-UN matrix with Cr amounts of 5 wt% and 10 wt%. The SPS densified Cr-UN composite pellets display greatly-improved thermal conductivity and simultaneously high hardness and fracture toughness. The fracture toughness of the Cr-incorporated UN pellet is ∼5.5 MPa-m1/2, representing almost 130% enhancement as compared to that of monolithic UN sintered at the same condition. The Cr-incorporated UN pellets also display enhanced oxidation resistance as evidenced by increased onset temperature for oxidation to 450 oC for the 5 wt% Cr-UN. These results highlight that Cr additive and the formation of a ternary phase can be useful to improve the thermal-mechanical properties and oxidation resistance of UN fuels with well-maintained high fissile element density.

Study on specific heat capacity and thermal conductivity of uranium nitride

Abstract In this study, we have proposed an analytical method for calculating the specific heat capacity of uranium nitride nuclear material. The specific heat capacity results have obtained by the use of the Debye-Einstein approximation. The thermal conductivity of nuclear material has been obtained by using the experimental data of thermal diffusivity and the calculation results of specific heat capacity. This method shows that our results are satisfactory for the wide range temperature variations. The proposed approach can be easily applied to determine the thermodynamic properties of the other nuclear materials.

Uranium nitride (UN) pellets with controllable microstructure and phase – fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties

Dense uranium mononitride (UN) pellets with controlled microstructures and tailored grain size from large-grained to a few microns are synthesized by spark plasma sintering (SPS) combined with high energy ball milling . The impacts of the sintering conditions on fuel microstructure, grain size, physical density, and phase behavior are systematically investigated, and the thermal-mechanical properties and oxidation behavior of the SPS densified UN pellets are characterized. Higher sintering temperatures and longer ball milling durations and thus finer starting UN powders promote sintering and densification , and dense UN pellets above 95% theoretical density can be achieved by SPS at 1873 K for 10 min. UN phase purity is maintained in the SPS-densified pellets sintered at a lower temperature and short duration. A phase heterogeneity with secondary UO 2 or uranium sesquinitride (U 2 N 3 ) occurs for the UN pellets sintered at higher temperatures using finer UN powders. The hardness and fracture toughness of the SPS-densified UN pellets increase with smaller grain sizes and higher densities to 7.9 GPa and 3.5 MPa m 1/2 , respectively. Both small (1–2 μm) and large grain-sized (30–50 µm) UN pellets exhibit good thermal conductivity . Dynamic oxidation testing by a thermogravimetric analyzer in air shows that the onset temperature for oxidation varies with microstructure and phase heterogeneity of the SPS densified UN pellets. Particularly, the smaller-grained (micron-sized) UN pellets containing uranium oxides and U 2 N 3 display lower weight gain and significantly-reduced oxidation kinetics , and full oxidation completes at a temperature above 1173 K when tested with a ramp rate of 10 K/min.

Uranium nitride advanced fuel: an evaluation of the oxidation resistance of coated and doped grains

The oxidation behaviour of the composite UN-AlN, UN-Cr 2 N/CrN and UN-AlN-Cr 2 N/CrN pellets in air and anoxic steam under thermal transient conditions was investigated and compared with the pure UN pellet. The composite pellets were manufactured to contain the engineered microstructure of coated (the addition of matrix-insoluble AlN) and doped (the addition of matrix-soluble Cr 2 N/CrN) grains. The composite powders were produced by powder metallurgy and sintered into pellets using the SPS method. Sintered composite pellets were subjected to a thermal transient up to 1273 K in an STA-EGA (TGA-DSC-Gas-MS) system, followed by crystallographic characterization by XRD and morphological and elemental analysis by FEG-SEM. Improved oxidation behaviour in air compared to pure UN was demonstrated by the UN-Cr 2 N/CrN composite pellet. The formation of the ternary oxide UCrO 4 from the ternary (U 2 Cr)N 3 phase (doped grain) was observed, consistent with the delayed oxidation onset and slower reaction rates. In an anoxic steam environment UN-Cr 2 N/CrN exhibited a higher onset oxidation temperature relative to UN, although the reaction progressed faster than for UN sample. Composite UN-AlN pellet oxidised at a lower temperature in both air and steam, compared to pure UN, due to internal stresses in the fuel matrix. A mechanism for degradation of the composite materials is proposed and the influence of the individual phases on the oxidation behaviour of the composites is discussed.

Digestion and trace metal analysis of uranium nitride

Digestion and trace metal analysis of uranium nitride

Oxidation and hydrolysis of thorium doped uranium nitride fuel for use in LWR

• Influence of thorium content on oxidation resistance of UN was observed • Thorium improve the oxidation resistance of low porosity UN pellets in air • Uranium nitride pellets sintered in SPS can withstand the interaction with water at 100°C • Loss of UN pellet integrity observed at 200°C, regardless of doping • Phase separation of ThO 2 and UO 2 was measured after water exposure experiment Uranium nitride is being investigated as a replacement for UO 2 as it shows enhanced thermal properties and seems to be a promising accident tolerant fuel (ATF) candidate. The main drawback of UN fuel is its innate low oxidation resistance in air/water environments. This becomes a challenge for the implementation of UN fuel in water-cooled reactors. The effect of thorium doping in the stability of uranium nitride microspheres and pellets sintered by spark plasma sintering (SPS) was investigated in oxidizing environments using thermogravimetric analysis and autoclave testing. It was found that during oxidation in air the density had a noticeable effect, increasing the reaction onset temperatures in pellets with higher densities. In addition, thorium doping improved the oxidation resistance of pellets in air by increasing the maximal reaction rate temperature by approximately 50 K. However, this effect was almost nonexistent in highly porous doped microspheres. The interaction with water at 373 K showed that pellets manufactured using SPS can survive unchanged for at least six hours in boiling water, which is an improvement to cold-pressed pellets. At 473 and 573 K, the pellets were oxidized and disintegration into an oxide powder was observed. Thorium-doped uranium nitride pellets did not present any improvement with respect to the oxidation resistance of UN in water at these temperatures.

Improvement of the BISON U3Si2 modeling capabilities based on multiscale developments to modeling fission gas behavior

• The recommended models for analyzing U 3 Si 2 fuel behavior with BISON are provided. • Lower-length scale methods have been used to reduce uncertainty in fission gas modeling of U 3 Si 2 . • Comparisons to PIE measurements of U 3 Si 2 fueled rodlets are acceptable. 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).

Diffusion in hypo-stoichiometric uranium mononitride

Uranium nitride is a nuclear fuel of interest as it offers enhanced accident tolerance, owing to its intrinsic properties. Before UN can be deployed commercially it is essential to understand its properties and how they evolve during operation. Therefore, molecular dynamics has been employed to study the thermal expansion and diffusivity of the intrinsic species in UN and how these change with stoichiometry. The introduction of hypostoichiometry as either nitrogen antisites or nitrogen vacancy defects is predicted to lead to an increase in nitrogen diffusivity and a concomitant decrease in the activation energy. The activation energies predicted for nitrogen diffusion in hypostoichiometry samples containing antisite defects are shown to offer a very close agreement with experimental observations.

Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1\u2212x)N Fuel

Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (ThxU1−x)N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between 0.05 le x le 0.5 have been considered for calculation of k_{infty } and k_{{{text{effective}}}}. ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

Uranium nitride triple bond is surprisingly covalent

An unusual uranium nitride complex contains a highly covalent triple bond that pushes the boundaries of actinide bonding ( Nat. Commun. 2021, DOI: 10.1038/s41467-021-25863-2 ). Covalency describes the degree of electron sharing between two atoms. In general, the earlier actinides tend to engage in ionic or polarized covalent bonding. This bonding character falls between the lanthanides, which prefer ionic bonding, and transition-metal complexes, which can have much more covalent bonds. But in 2013, a team led by Stephen T. Liddle, now at the University of Manchester, created a terminal uranium(VI)-nitride complex (shown) that was computationally predicted to have a remarkably covalent U≡N bond. Liddle’s team has now experimentally confirmed this with 15 N nuclear magnetic resonance spectroscopy, after synthesizing a version of the complex containing the isotope. NMR data showed that the U≡N bond is even more covalent than analogous transition metal–nitride complexes. “We were really surprised,” Liddle says. “We

Uranium nitride triple bond is surprisingly covalent

Uranium nitride triple bond is surprisingly covalent