| 1. |
- Jerndal, Erik, 1980, et al.
(författare)
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Using Low-Cost Iron-Based Materials as Oxygen Carriers for Chemical Looping Combustion
- 2011
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Ingår i: Oil and Gas Science and Technology. - : EDP Sciences. - 1294-4475 .- 1953-8189. ; 66:2, s. 235-248
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Tidskriftsartikel (refereegranskat)abstract
- In chemical looping combustion with solid fuels, the oxygen-carrier lifetime is expectedto be shorter than with gaseous fuels. Therefore, it is particularly important to use low-cost oxygencarriers in solid fuel applications. Apart from being cheap, these oxygen carriers should be able toconvert the CO and H2 produced from the solid fuel gasification and be sufficiently hard to withstandfragmentation. Several low-cost iron-based materials displayed high conversion of syngas and highmechanical strength and can be used for further development of the technology. These materials includeoxide scales from Sandvik and Scana and an iron ore from LKAB. All tested oxygen carriers showedhigher gas conversion than a reference sample, the mineral ilmenite. Generally, softer oxygen carrierswere more porous and appeared to have a higher reactivity towards syngas. When compared withilmenite, the conversion of CO was higher for all oxygen carriers and the conversion of H2 was higherwhen tested for longer reduction times. The oxygen carrier Sandvik 2 displayed the highest conversion ofsyngas and was therefore selected for solid fuel experiments. The conversion rate of solid fuels washigher with Sandvik 2 than with the reference sample, ilmenite.
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| 2. |
- Tiberiu Costin, Dan, 1983, et al.
(författare)
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Thermochemical effect of fission products on sodium – MOX fuel reaction: The case of niobium
- 2018
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 500, s. 361-365
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 3. |
- Aneheim, Emma, 1982, et al.
(författare)
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Dissolution Performance of Plutonium Nitride Based Fuel Materials
- 2016
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Ingår i: Atalante 2016 International Conference on Nuclear Chemistry for Sustainable Fuel Cycles. - : Elsevier BV. - 1876-6196. ; 21, s. 231-238
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Konferensbidrag (refereegranskat)abstract
- 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.
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| 4. |
- Bauhn, Lovisa, 1981, et al.
(författare)
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A micro hot test of the Chalmers-GANEX extraction system on used nuclear fuel
- 2013
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Ingår i: International Nuclear Fuel Cycle Conference, GLOBAL 2013: Nuclear Energy at a Crossroads. ; 1, s. 335-340
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Konferensbidrag (refereegranskat)abstract
- In the present study, a "micro hot test" has been performed using the Chalmers-GANEX (Group ActiNide EXtraction) system for partitioning of used nuclear fuel. The test included a pre-extraction step using N,N-di-2- ethylhexylbutyramide (DEHBA) in n-octanol to remove the bulk part of the uranium. This pre-extraction was followed by a group extraction of actinides using the mixture of TBP and CyMe4-BTBP in cyclohexanone as suggested in the Chalmers-GANEX process, and a three stage stripping of the extracted actinides. Distribution ratios for the extractions and stripping were determined based on a combination of γ- and α-spectrometry, as well as ICP-MS measurements. Successful extraction of uranium, plutonium and the minor actinides neptunium, americium and curium was achieved. However, measurements also indicated that co-extraction of europium occurs to some extent during the separation. These results were expected based on previous experiments using trace concentrations of actinides and lanthanides. Since this test was only performed in one stage with respect to the group actinide extraction, it is expected that multi stage tests will give even better results.
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| 5. |
- Costa, Diogo Ribeiro, et al.
(författare)
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Coated ZrN sphere-UO2 composites as surrogates for UN-UO2 accident tolerant fuels
- 2022
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 567, s. 153845-
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 6. |
- Costa, Diogo Ribeiro, et al.
(författare)
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Oxidation of UN/U 2 N 3 -UO 2 composites: an evaluation of UO 2 as an oxidation barrier for the nitride phases
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 544
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 7. |
- Costa, Diogo Ribeiro, et al.
(författare)
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UN microspheres embedded in UO2 matrix: An innovative accident tolerant fuel
- 2020
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 540
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 8. |
- Ekberg, Christian, 1967, et al.
(författare)
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Nitride fuel for Gen IV nuclear power systems
- 2018
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Ingår i: Journal of Radioanalytical and Nuclear Chemistry. - : Springer Science and Business Media LLC. - 0236-5731 .- 1588-2780. ; 318:3, s. 1713-1725
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Forskningsöversikt (refereegranskat)abstract
- 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.
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| 9. |
- Gonzalez Fonseca, Luis Guillermo, 1993, et al.
(författare)
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Application of SPS in the fabrication of UN and (U,Th)N pellets from microspheres
- 2020
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 536
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Tidskriftsartikel (refereegranskat)abstract
- In this study, the process involved in the fabrication of a potential accident tolerant fuel is described. Homogeneous uranium nitride microspheres doped with different thorium content were successfully manufactured using an internal gelation process followed by carbothermic reduction, and nitridation. Elemental analysis of the materials showed low carbon and oxygen content, the two major impurities found in the products of carbothermic reduction. Uranium nitride microspheres were pressed and sintered using spark plasma sintering (SPS) to produce pellets with variable density. Final density can be tailored by choosing the sintering temperature, pressure and time. Density values of 77–98% of theoretical density (%TD) were found. As expected, higher temperatures and pressures resulted in a denser material. Furthermore, a direct correlation between the onset sintering temperature and thorium content in the materials was observed. The change of onset temperature has been related to an increment in the activation energy for self-diffusion due to the substitution of uranium atoms by thorium in the crystal structure.
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| 10. |
- Gonzalez Fonseca, Luis Guillermo, 1993, et al.
(författare)
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Oxidation and hydrolysis of thorium doped uranium nitride fuel for use in LWR
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 555
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Tidskriftsartikel (refereegranskat)abstract
- Uranium nitride is being investigated as a replacement for UO2 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.
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