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 mu m and 65 mu m, respectively, while the binder and CVD of W methods layers of about 12 mu m and 3 mu m, respectively. (C) 2022 The Author(s). Published by Elsevier B.V.

Composite fuels such as UN-UO2 are being considered to address the lower oxidation resistance of the UN fuel from a safety perspective for use in light water reactors, whilst improving the in-reactor behaviour of the more ubiquitous UO2 fuel. An innovative UN-UO2 accident tolerant fuel has recently been fabricated and studied: UN microspheres embedded in UO2 matrix. In the present study, detailed oxidative thermogravimetric investigations (TGA/DSC) of high-density UN/U2N3-UO2 composite fuels (91-97 %TD), as well as post oxidised microstructures obtained by SEM, are reported and analysed. Triplicate TGA measurements of each specimen were carried out at 5 K/min up to 973 K in a synthetic air atmosphere to assess their oxidation kinetics. The mass variation due to the oxidation reactions (%), the oxidation onset temperatures (OOTs), and the maximum reaction temperatures (MRTs) are also presented and discussed. The results show that all composites have similar post oxidised microstructures with mostly intergranular cracking and spalling. The oxidation resistance of the pellet with initially 10 wt% of UN microspheres is surprisingly better than the UO2 reference. Moreover, there is no significant difference in the OOT (~557 K) and MRT (~615 K) when 30 wt% or 50 wt% of embedded UN microspheres are used. Therefore, the findings in this article demonstrate that the UO2 matrix acts as a barrier to improve the oxidation resistance of the nitride phases at the beginning of life conditions.

A comparative study of chemical looping combustion (CLC) with paper, plastic, and coal as fuel was carried out. Experiments were performed in a laboratory fluidized-bed reactor by alternating between reduction and oxidation cycles. The results obtained indicated that a higher temperature leads to an increase in the CO2 yield and carbon conversion for all fuels. Paper had the highest fractional conversion of CO to CO2 followed by polyvinyl chloride (PVC) and coal. This was due to the higher fraction of volatiles in paper compared to PVC and coal. Scanning electron microscopy (SEM) analysis of the oxygen carrier particle after each of the solid fuel experiment was carried out. For the used ilmenite, there was a slight difference in the morphology for the three different fuels.

Uranium nitride (UN)-uranium dioxide (UO2) composite fuels are being considered as an accident tolerant fuel (ATF) option for light water reactors. However, the complexity related to the chemical interactions between UN and UO(2 )during sintering is still an open problem. Moreover, there is a lack of knowledge regarding the influence of the sintering parameters on the amount and morphology of the alpha-U2N3 phase formed. In this study, a detailed investigation of the interaction between UN and UO2 is provided and a formation mechanism for the resulting alpha-U2N3 phase is proposed. Coupled with these analyses, an innovative ATF concept was investigated: UN microspheres and UO2,13 powder were mixed and subsequently sintered by spark plasma sintering. Different temperatures, pressures, times and cooling rates were evaluated. The pellets were characterised by complementary techniques, including XRD, DSC, and SEM-EDS/WDS/EBSD. The UN and UO2 interaction is driven by O diffusion into the UN phase and N diffusion in the opposite direction, forming a long-range solid solution in the UO2 matrix, that can be described as UO2-xNx. The cooling process decreases the N solubility in UO2-xNx, causing then N redistribution and precipitation as alpha-U2N3 phase along and inside the UO2 grains. This precipitation mechanism occurs at temperatures between 1273 K and 973 K on cooling, following specific crystallographic grain orientation patterns. (C) 2020 The Authors. Published by Elsevier B.V.

Combustion of organic materials in fluidized bed combustion is generally performed using an over‐stoichiometric air‐to‐fuel ratio. Despite excess air in the system, sub‐stoichiometric combustion regions are present in the fluidized bed because of nonperfect mixing of reactants in the system. These regions with oxygen‐deficient combustion contribute to increased levels of nonoxidized or partially oxidized carbon species. To enhance heat transfer and uniform heat distribution in the incineration chamber, an inert fluidized bed material such as silica sand is generally used. Substitution of silica sand in favor of an oxygen carrier could potentially be used to promote oxygen distribution in the incineration chamber. This is referred to as oxygen carrier‐aided combustion. In this work, three alternative bed materials, previously investigated in chemical looping combustion, are investigated and compared with silica sand in over‐ and sub‐stoichiometric combustion. The materials investigated as bed materials are a manganese ore, the mineral ilmenite, and a synthetic material mixture of Fe2O3 on a ZrO2 support. Results show that during combustion using sub‐stoichiometric air‐to‐fuel ratios, the amount of CO in the effluent gases can be reduced using an active bed material compared to inert silica sand.

Uranium nitride (UN) is considered as nuclear reactor fuel because of, among other reasons, its high uranium density and its high thermal conductivity. Its main drawback is that it relatively easily dissolves in hot water, which is particularly problematic when it is used in water-cooled reactors. One possible remedy to this is to add some corrosion inhibitor as dopant to the UN matrix. A number of dopants have been identified that have the potential to inhibit the dissolution process, and their respective merits have been investigated both by neutronic simulations and dissolution experiments. It is concluded that chromium is the most promising candidate.

Nuclear energy has been a part of the energy mix in many countries for decades. Today in principle all power producing reactors use the same techniqe. Either PWR or BWR fuelled with oxide fuels. This choice of fuel is not self evident and today there are suggestions to change to fuels which may be safer and more economical and also used in e.g. Gen IV nuclear power systems. One such fuel type is the nitrides. The nitrides have a better thermal conductivity than the oxides and a similar melting point and are thus have larger safety margins to melting during operation. In addition they are between 30 and 40% more dense with respect to fissile material. Drawbacks include instability with respect to water and a sometimes complicated fabrication route. The former is not really an issue with Gen IV systems but for use in the present fleet. In this paper we discuss both production and recycling potential of nitride fuels.

The influence of niobium on the sodium MOX fuel chemical interaction was studied by different heat treatments of airtight capsules containing fresh MOX, sodium and a niobium strip. The characterisation results evidenced a two-step process with first MOX oxidation and then MOX reduction. This result was interpreted by considering the formation of sodium niobiate that captures oxygen from the MOX. This interpretation is used to discuss the influence of niobium as fission product on the sodium –irradiated MOX fuel reaction.

Sol-gel based techniques are one way to lower the handling of highly radioactive powders when producing transuranium-containing fuel. In this work inert matrix (Zr0.6,Pu0.4)N fuel has been produced by internal gelation followed by carbothermal reduction. No co-gelation was observed during internal gelation and a two phase material could be detected by scanning electron microscopy in the nitrided microspheres. Sintering has been performed in both Ar and N2. X-ray diffraction revealed that sintering in N2 produced a solid solution, while sintering in Ar did not. The final metal composition in the microspheres was determined by ICP-MS to be about 41% Pu and 59% Zr. Vegard's law was applied to estimate the nitride purity in the solid solution pellet to be Zr0.6Pu0.4N0.87C0.13 making the final material more of a carbonitride than a pure nitride.

Fabrication of plutonium containing fuels through the internal gelation method has mostly been studied in mixed metal systems such as U, Pu or Zr, Pu. In this work production of undiluted PuN has been performed by carbothermal reduction on both oxide powder and Pu microspheres produced by the internal gelation method. Nitride purities reached using the different methods have been studied together with final densities achieved during pellet fabrication as well as losses of ingrown Am during the different production steps. Formation of Pu microspheres was successfully performed using the internal gelation method, although extensive microsphere fracturing occurred during thermal treatment. Final densities of PuN pellets produced by cold pressing and sintering reached 70-80% of theoretical density. Am losses during the carbothermal reduction step was on average about 3.7%. After sintering about 11% of Am was lost in total through the entire production process if sintering in N-2 + 5% H-2 atmosphere while about 50% of the Am in total was lost when using Ar as sintering atmosphere. (C) 2016 Elsevier B.V. All rights reserved.

Inert matrix fuel is a fuel type where the fissile material is blended with a solid diluent material. In this work zirconium carbonitride microspheres have been produced by internal sol gel technique, followed by carbothermal reduction. Material nitride purities in the produced materials ranged from Zr(N0.45C0.55) to Zr(N0.74C0.26) as determined by X-ray diffraction and application of Vegard's law. The zirconium carbonitride microspheres have been pelletized by spark plasma sintering (SPS) and by conventional cold pressing and sintering. In all SPS experiments cohesive pellets were formed. Maximum final density reached by SPS at 1700 degrees C was 87% theoretical density (TD) compared to 53% TD in conventional sintering at 1700 degrees C. Pore sizes in all the produced pellets were in the mu m scale and no density gradients could be observed by computer tomography. (C) 2016 Elsevier B.V. All rights reserved.

Nitride fuels have been regarded as one viable fuel option for Generation IV reactors due to their positive features compared to oxides. To be able to close the fuel cycle and follow the Generation IV concept, nitrides must, however, demonstrate their ability to be reprocessed. This means that the dissolution performance of actinide based nitrides has to be thoroughly investigated and assessed. As the zirconium stabilized nitrides show even better potential as fuel material than does the pure actinide containing nitrides, investigations on the dissolution behavior of both PuN and (Pu,Zr)N has been undertaken. If possible it is desirable to perform the fuel dissolutions using nitric acid. This, as most reprocessing strategies using solvent-solvent extraction are based on a nitride containing aqueous matrix.(Pu,Zr)N/C microspheres were produced using internal gelation. The spheres dissolution performance was investigated using nitric acid with and without additions of HF and Ag(II). In addition PuN fuel pellets were produced from powder and their dissolution performance were also assessed in a nitric acid based setting. (C) 2016 The Authors. Published by Elsevier B.V.

In chemical looping combustion with solid fuels, the oxygen-carrier lifetime is expected to be shorter than with gaseous fuels. Therefore, it is particularly important to use low-cost oxygen carriers in solid fuel applications. Apart from being cheap, these oxygen carriers should be able to convert the CO and H2 produced from the solid fuel gasification and be sufficiently hard to withstand fragmentation. Several low-cost iron-based materials displayed high conversion of syngas and high mechanical strength and can be used for further development of the technology. These materials include oxide scales from Sandvik and Scana and an iron ore from LKAB. All tested oxygen carriers showed higher gas conversion than a reference sample, the mineral ilmenite. Generally, softer oxygen carriers were more porous and appeared to have a higher reactivity towards syngas. When compared with ilmenite, the conversion of CO was higher for all oxygen carriers and the conversion of H2 was higher when tested for longer reduction times. The oxygen carrier Sandvik 2 displayed the highest conversion of syngas and was therefore selected for solid fuel experiments. The conversion rate of solid fuels was higher with Sandvik 2 than with the reference sample, ilmenite.