Investigation of ThO2 and (U, Th)O2

The effect of the properties of ThO₂ and (U, Th)O₂ powders, prepared with different technological regimes, on the properties of the finished items is investigated. The work includes detailed investigations of ThO₂ and (U, Th)O₂ powders (x-ray phase analysis, electron-microscope investigation) and sintered fuel pellets (determination of density, study of microstructure, thermophysical investigations). The temperature dependences of the crystal lattice parameters and the sizes of the crystallites in ThO₂ and (U, Th)O₂ powders with different UO₂:ThO₂ ratio are obtained. The temperature dependences of the thermal conductivity of sintered ThO₂ and (U, Th)O₂ pellets with different UO₂:ThO₂ ratio are studied.     

Effect of the heat-treatment temperature of powders on oxide fuel quality

It has been determined at the State Scientific Center of the Russian Federation—Physics and Power Engineering Institute in the course of developing a technology for fabricating various fuel compositions (UO₂+MgO, UO₂+ThO₂, UO₂+Th+ThO₂, PuO₂+MgO, UO₂+Fe+MgO, PuO₂+BaO, and others) for fast-neutron and light-water reactors that structural changes in particle aggolmerates occur at the heat-treatment stage. The optimal properties of the powders are obtained at the temperature of the morphological transformations of the particles. The fuel pellets prepared from these powders possess stable density, porosity, exterior form, mechanical strength, and so on. The total specific surface area of the oxides is an indirect parameter for estimating their quality. Each fuel composition has its own optimal powder heat-temperature temperature. 7 figures, 1 table, 5 references. State Scientific Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 189–194, March, 1999.     

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,²³³U)O₂ mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with ²³²U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO₂ is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O₂ pellets. In this study, fabrication of (Th,U)O₂ mixed oxide pellets containing 3–5 wt.% UO₂ was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.     

Phase studies in the UO2-Gd2O3 system

The incorporation of gadolinium directly into nuclear fuel is important regarding reactivity compensation, which enables longer fuel cycles. The incorporation of Gd₂O₃ powder directly into the UO₂ powder by dry mechanical blending is the most attractive process, because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. This is due to the bad sintering behavior of the UO₂-Gd₂O₃ mixed fuel, which shows a blockage in the sintering process that hinder the densification process. Minimal information exists regarding the possible mechanisms for this blockage and this is restricted to the hypothesis based on the formation of a low diffusivity Gd rich (U,Gd)O₂ phase. The objective of this investigation was to study the phase formation in this system, thus contributing to clarifying the causes of the blockage. Experimental evidence indicated the existence of phases in the (U,Gd)O₂ system that revealed structures different from the fluorite-type UO₂ structure. These phases appear to be isostructural to the phases observed in the rare earth-oxygen system.     

The effect of processing on the thermal diffusivity of MgO-Nd2Zr2O7 composites for inert matrix materials

Oxides possess many of the required properties suitable for an inert matrix fuel in light water reactors, however, their primary disadvantage is low thermal conductivity. Composites are being investigated to maximize the thermal conductivity of the inert matrix fuel by using thermally conductive MgO as the primary phase while improving its hot water corrosion resistance through the addition of a second phase acting as a hydration barrier. Inert matrix fuel candidate MgO-Nd₂Zr₂O₇ composites were synthesized with multiple processing methods, the composite powders were characterized, the resulting microstructures quantitatively analyzed, and the thermal diffusivity of the composites was measured. Among the four processing methods investigated, ball milling and high-energy shaker blending produced the most homogeneous microstructures with a negligible amount of MgO and Nd₂Zr₂O₇ heterogeneities. An effect of processing on the properties of the composites manifests as a larger variation in the thermal diffusivity in pellets processed by methods that produce a higher quantity and frequency of MgO and Nd₂Zr₂O₇ heterogeneities than in methods that produce negligible amounts of heterogeneities.     

Characterization of ThO2-UO2 pellets made by co-precipitation process

The co-precipitation technique renders an excellent route to obtain a homogeneous mixture of ThO₂ and UO₂ powders. In this process, after the nitrate solutions of Th and U are mixed in the intended ratio, oxalic acid is added for co-precipitation. The precipitate is then dried and calcined to get a solid solution of ThO₂ and UO₂. In this study, ThO₂-30%UO₂ and ThO₂-50%UO₂ (% in weight) powders were characterized in terms of particle size, particle shape, surface area, phase content, O/M ratio etc. The pellets obtained by sintering these powders were characterized with the help of optical microscopy, scanning electron microscopy (SEM) and electron probe microanalysis (EPMA). The XRD data for ThO₂-30%UO₂ and ThO₂-50%UO₂ pellets showed the presence of a small amount of U₃O₈ phase besides fluorite phase. The grain size of ThO₂-30%UO₂ and ThO₂-50%UO₂ was found to be 5.7 and 4.5 μm, respectively.     

Mössbauer study and structural characterization of UO2–Gd2O3 sintered compounds

Samples of UO₂and up to 10 wt% of Gd₂O₃ were prepared by solid-state reaction under a reducing atmosphere, in a thermal path comprising ramps and dwell times in the temperature range of 900–1750 °C. The sintered material was analyzed by X-ray diffraction and ¹⁵⁵Gd Mössbauer spectroscopy. The results showed that for samples annealed up to 900 °C, the gadolinium sesquioxide remained unreacted. However, when the temperature was increased to 1300 °C, a solid-state reaction took place forming mixed oxides. For the more severe sintering condition, at 1750 °C, gadolinia left urania partially unreacted producing a material consisting of two compositions, UO₂ (with no dissolved gadolinium) and (U, Gd)O₂. The proposed heating cycle provided pellets free from Gd₂O₃ phase and may be used by the nuclear fuel industry as a suitable sintering process.     

Treatment of UO2 pellets used for preparing fuel elements of HTR-10 followed by supercritical fluid extraction

International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO₂ pellets used for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) followed by supercritical fluid extraction was investigated. When UO₂ pellets were dissolved and extracted with tri-n-butyl phosphate (TBP)–HNO₃ complex in supercritical CO₂ (SC-CO₂), the extraction efficiency was less than 7% under experimental conditions. After UO₂ pellets were ground into UO₂ fine powders, the extraction efficiency of the UO₂ fine powders with TBP–HNO₃ complex in SC-CO₂ could reach 92%. After UO₂ pellets broke spontaneously into U₃O₈ powders under O₂ flow and 600 °C, the extraction efficiency of the U₃O₈ powder with TBP–HNO₃ complex in SC-CO₂ could reach more than 98%.     

Improvement of UO2 Pellet Properties by Controlling the Powder Morphology of Recycled U3O8 Powder

Powder morphology evolution of recycled U₃O₈ according to the thermal treatments has been studied. The defective UO₂ pellets are oxidized to U₃O₈ powders at a conventional temperature of 350 or 450°C in air. Those powders are pressed into green pellets and then sintered at 1,500 and 1,730°C in H₂ gas flow. Final reoxidized U₃O₈ powers are obtained by reoxidizing those sintered pellets at 450°C in air. This paper shows that the reoxidized U₃O₈ powder morphology and the BET surface areas are greatly dependent on the density of sintered UO₂ pellets before reoxidation. Reoxidized U₃O₈ powders are added to virgin UO₂ powders to fabricate UO₂ pellets and the effect of such addition on the UO₂ pellet properties is investigated. The reoxidized U₃O₈ powders having a certain range of BET surface area significantly promote the grain growth of UO₂ pellets.     

Fabrication and study of the properties of PuO2−MgO fuel pellets

Conclusions Defect-free PuO₂−MgO pellets with a density of 4.4 g/cm³ (90% of the computed density of the composition, in which the volume fractions of PuO₂ and MgO equal 15 and 85% respectively), were obtained. Work with plutonium-containing material showed that the technology developed for fabricating UO₂−MgO fuel pellets is suitable for fabricating PuO₂−MgO fuel pellets. Main Science Center of the Russian Federation — A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 82, No. 5, pp. 355–358, May, 1997.     

A mechanism for the sintered density decrease of UO2–Gd2O3 pellets under an oxidizing atmosphere

A mixture of UO₂ and Gd₂O₃ powders was pressed into compacts and sintered under various atmospheres ranging from reducing to oxidizing gases. The sintered density of UO₂–10 wt% Gd₂O₃ pellets decreases with increasing oxygen potential of the sintering atmosphere. Dilatometry and X-ray diffraction studies indicate that the delay of densification takes place between 1300°C and 1500°C, along with the formation of (U,Gd) O₂. A very large solubility of Gd₂O₃ in UO₂ relative to the reverse solubility might cause Gd ions to diffuse into UO₂ so directionally that new pores are produced at the places of Gd₂O₃ particles. The new pores may be difficult to shrink and thus lead to the density decrease under an oxidizing atmosphere but not under a reducing atmosphere, because a driving force for the shrinkage of new pores may be smaller under an oxidizing atmosphere than under a reducing atmosphere.     

Fabrication of particle dispersed inert matrix fuel based on liquid phase sintered SiC

In the present work, liquid phase sintered SiC (LPS-SiC) was proposed as an inert matrix for the particle dispersed inert matrix fuel (IMF). The fuel particles containing plutonium and minor actinides were substituted with pure yttria stabilized zirconia beads. The LPS-SiC matrix was produced from the initial mixtures prepared using submicron sized α-SiC powder and oxide additives Al₂O₃, Y₂O₃ in the amount of 10 wt.% with the molar ratio 1Y₂O₃/1Al₂O₃. Powder mixtures were sintered using two sintering methods; namely conventional high temperature sintering and novel spark plasma sintering at different temperatures depending on the method applied in order to obtain dense samples. The phase reaction products were identified using X-ray diffraction (XRD) and microstructures were investigated using light microscopy and scanning electron microscopy with energy dispersive X-ray spectroscopy (SEM/EDX) techniques. The influence of powder mixing methods, sintering temperatures, pressures applied and holding time on the density of the obtained pellets was investigated. The samples sintered by slow conventional sintering show lower relative density and more pronounced interaction between the fuel particles and matrix in comparison with those obtained with the fast spark plasma sintering method.     

Irradiation behavior of experimental fuel particles containing chemically vapor deposited zirconium carbide coatings

Four experimental fuel particle designs, utilizing zirconium carbide coatings in combination with porous and dense pyrocarbon coatings, were tested under high-temperature irradiation. As a fission-product corrosion test for the zirconium carbide, two particle designs employed carbide coatings applied directly over either UC₂ or (8Th, 1U)O₂ fuel kernels. The other two designs utilized zirconium carbide outside of porous pyrocarbon coatings but without the conventional inner dense pyrocarbon coating on either UC₂ or (8Th, 1U)O₂ fuel kernels. The particles were irradiated at 1200°C to a fast-neutron fluence of $\text{5 × 10}{}^{21}\text{n/cm}{}^2\text{(E μ'0.18 MeV)}$ and fractional burnups of the initial metal atoms of 0.7 and 0.08 for the UC₂ and (8Th, 1U)O₂ kernels, respectively. Stereoscopic, metallographic and electron-beam microprobe examination of the irradiated particles showed that the zirconium carbide possesses exceptional resistance to chemical attack by fission products and good mechanical stability under irradiation.     

A practical fabrication method for the advanced heterogeneous fuel with magnesia containing minor actinides

Fabrication tests on advanced heterogeneous fuel with MgO were carried out for the purpose of establishing a practical fabrication method. Advanced heterogeneous fuel consists of spheres (diameter greater than 100 μm) of a minor actinide oxide and MgO inert matrix (macro-dispersed type fuel). Macro-dispersed type fuel pellets with a high density above 90% T.D. were successfully fabricated. In addition, the fabricated pellets showed a homogeneous dispersion of near spherical host phase granules. These were attained by optimization of the fabrication process and conditions; i.e. a preliminary heat treatment of raw powders of host phase, an adjustment of pressure at the granulation process, deployment of a spray-dry process for MgO sphere preparation and sintering in a He atmosphere. From these results, a practical fabrication method for MgO-based macro-dispersed type fuel based on a simple powder metallurgical technique was established.     

Preparation of High Density Uranium Nitride and Uranium Carbonitride Fuel Pellets

Uranium nitride and uranium carbonitride fuel pellets were prepared for irradiation in the Japan Material Testing Reactor. The pellets are 6.9 mm in diameter and 7 mm long, and are of natural and 5% enriched uranium. Uranium nitride powder was prepared from uranium metal via hydride and higher nitride. Uranium carbide powder was prepared from uranium metal by hydriding and then reacting with propane. The lowest possible reaction temperatures were selected to obtain fine and reactive powders. Uranium nitride and mixed powders of different ratios (UC: UN = 1: 3, 1:1 and 3: 1) were cold pressed without binder. Sintering was carried out in a tungsten crucible in vacuum (10~⁴ mmHg) for 2 hr at 1,900°–2,000°C. The density of the pellets obtained was in the range of 90~95% of the theoretical value with an oxygen content of 1,300~2,100 ppm. No second phase, such as metallic uranium, were observed in the specimens, either by metallography or X-ray diffraction. These pellets of unexpectedly high density without second phase must have been obtained thanks to the good powder characteristics combined with proper sintering conditions. The compositions of uranium carbonitride pellets were found to be slightly nitrogen deficient, compared with the reactants.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

(U,Gd)O_2可燃毒物燃料芯块制造工艺及其主要性能

介绍了中国核动力研究设计院(U,Gd)O2可燃毒物燃料芯块制造生产线所使用的原材料UO2粉末、Gd2O3粉末、(U,Gd)3O8粉末以及添加剂硬脂酸锌和草酸铵的主要性能,同时描述了混料、制粒、成型、烧结和磨削等制造工艺过程及其产品(U,Gd)O2芯块的主要性能,并对制造过程中有关工艺控制参数进行了讨论。     

Study of the properties of ammonium polyuranate precipitates and UO<Subscript>2</Subscript> powders and pellets obtained with different technologies

The results of a differential-thermal analysis are used to compare the properties of ammonia polyuranate precipitates, UO₂ powders and pellets, obtained by different methods as well as metallic uranium. It is found that the phase NH₃·3UO₃·5H₂O forms in regular precipitation of ammonium polyuranate. When using nanotechnology, the phases NH₃·2UO₃·3H₂O and 4NH₃·6UO₃·8H₂O are also present in the precipitate. UO₂ powder prepared from such precipitate has high activity, since all phase transformations in it occur at a lower temperature. Modified fuel pellets of uranium dioxide, which are obtained by means of nanotechnology or mechanical addition of ammonia-containing reagents to powder, differ from the standard powders by a lower rate and more complex mechanism of oxidation, similar to metallic uranium. Modified UO₂ fuel pellets fabricated at the Physics and Power-Engineering Institute, are now undergoing tests in the BOR-60 reactor. After tests on the irradiated new modified fuel have been completed, it will be possible to judge its reliability.     

Potential nuclear fuels in the US-UC-UC2 system

A solid-solution study of the US-UC-UC₂ system at 1760 and 2170 K has revealed the range of compositions yielding the mandatory single-phase solid solutions for consideration as potential nuclear fuels. Two such compositions containing 20 and 40 mol%, respectively, of dissolved carbides have been selected, and convenient methods are described for their preparation from stoichiometric UC, ZnS and U₃O₈. Many of the physical and chemical properties of these compositions either as powders or as sintered pellets are reported. Compatibility with stainless steels under the test conditions is extremely good. However, sodium bonding is precluded through chemical reactions leading to the formation of Na₂S, and ultimately free uranium. Creep strengths are greater than for hyperstoichiometric UC. It is concluded that the 80 mol% US composition with less than 10 mol% UC₂is promising as a potential nuclear fuel, the 60 mol% US material being less so.     

Influence of O/M ratio on sintering behavior of (U0.8, Pu0.2)O2±x

The shrinkage of (U_0.8, Pu_0.2)O_2±x pellets was investigated with the help of a thermal dilatometer in isothermal and isochronal heating tests. During shrinkage measurements in isothermal heating, the oxygen-to-metal ratio of the pellets was maintained at a constant value by controlling the oxygen potential in the sintering atmosphere. The influence of the oxygen-to-metal ratio on the sintering behavior was evaluated from the measurement results. Mainly two mechanisms dominated the sintering of mixed oxide pellets. When the oxygen-to-metal ratio was close to the stoichiometric composition, pellet shrinkage progressed at low temperatures of 1200-1600 K, and the shrinkage rate of the pellets drastically changed with a small deviation from stoichiometric composition. The result showed that a diffusion process was dominated during the sintering of near-stoichiometric compositions. On the other hand, the sintering of reduced mixed oxide pellet proceeded at high temperatures of 1600-1900 K, and the shrinkage rate was very low as compared with stoichiometric mixed oxide.