An investigation on the green properties of U-10wt.%Mo/Al and U3Si2/Al powder compacts

The effects of fuel powder volume fraction and fuel particle shape on green properties of compacts, which were produced by processing the blended U-10wt.%Mo and U₃Si₂ with Al powders were investigated respectively, with respective to the compacting pressure range of 50–400 MPa. The relative density of the compacts increases with decreasing volume fraction of fuel powder. The compressibility of comminuted powder compacts was larger than that of the atomized powder compacts due to the fragmentation of comminuted particles, and the compressibility of the compacts of U-10wt.%Mo was larger than that of the compacts of U₃Si₂ due to the deformation of U-10wt.%Mo particles. The green strength of the comminuted powder compacts is higher than that of the atomized powder compact. This seems to have resulted from the smaller pore size and the larger contact area between the comminuted fuel powders and Al powders. It is suggested that the compacting condition adjustment be required to fabricate the atomized powder compacts having comparable green strength.     

The kinetics of U3Si formation in cast U3Si alloy

The growth of U₃Si in cast U-3.8 wt % Si alloy was measured by determining both the U₃Si₂/U₃Si interface movement, and the change in the total amount of the various phases. The activation energy controlling growth was found to be 50 ± 1 kcal/mole.     

Characterization of the interaction layer grown by interdiffusion between U-7wt%Mo and Al A356 alloy at 550 and 340 °C

U(Mo) alloys are under study to get a low-enriched U fuel for research and test reactors. Qualification experiments of dispersion fuel elements have shown that the interaction layer between the U(Mo) particles and the Al matrix behaves unsatisfactorily. The addition of Si to Al seems to be a good solution. The goal of this work is to identify the phases constituting the interaction layer for out-of-pile interdiffusion couples U(Mo)/Al(Si). Samples γU-7wt%Mo/Al A356 alloy (7.1 wt%Si) made by Friction Stir Welding were annealed at 550 and 340 °C. Results from metallography, microanalysis and X-ray diffraction, indicate that the interaction layer at 550 °C is formed by the phases U(Al,Si)₃, U₃Si₅ and Al₂₀Mo₂ U, while at 340 °C it is formed by U(Al,Si)₃ and U₃Si₅. X-ray diffraction with synchrotron radiation showed that the Si-rich phase, previously reported in the interaction layer at 550 °C near U(Mo) alloy, is U₃Si₅.     

The ternary system: silicon-uranium-vanadium

Phase equilibria in the system Si-U-V were established at 1100 °C by optical microscopy, EMPA and X-ray diffraction. Two ternary compounds were observed, U₂V₃Si₄ and (U_1−xV_x)₅Si₃, for which the crystal structures were elucidated by X-ray powder data refinement and found to be isotypic with the monoclinic U₂Mo₃Si₄-type (space group P2₁/c; a = 0.6821(3), b = 0.6820(4), c = 0.6735(3) nm, β = 109.77(1)°) and the tetragonal W₅Si₃-type (space group I4/mcm, a = 1.06825(2), c = 0.52764(2) nm), respectively. (U_1−xV_x)₅Si₃ appears at 1100 °C without any significant homogeneity region at x ∼ 0.2 resulting in a formula U₄VSi₃ which corresponds to a fully ordered atom arrangement. DTA experiments clearly show decomposition of this phase above 1206 °C revealing a two-phase region U₃Si₂ + V₃Si. At 1100 °C U₄VSi₃ is in equilibrium with V₃Si, V₅Si₃, U₃Si₂ and U(V). At 800 °C U₄VSi₃ forms one vertex of the tie-triangle to U₃Si and V₃Si. Due to the rather high thermodynamic stability of V₃Si and the corresponding tie-lines V₃Si + liquid at 1100 °C and V₃Si + U(V) below 925 °C, no compatibility exists between U₃Si or U₃Si₂ and vanadium metal.     

Post-irradiation examination of AlFeNi cladded U3Si2 fuel plates irradiated under severe conditions

Three full size AlFeNi cladded U₃Si₂ fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U₃Si₂ fuel and the Al matrix, has been quantified as U(Al,Si)_4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U₃Si₂ are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U₃Si₂ the bubbles are larger but there are fewer.     

Reactivity temperature coefficients for the HEU and LEU fuel of the MNSR reactor

Analysis of the Reactivity Temperature Coefficients of the Miniature Neutron Source Reactor (MNSR) for normal and accidental conditions (above 45 °C) using HEU-UAl₄ and the LEU: U₃Si, U₃Si₂ and U9Mo fuel were carried out in this paper. The Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and Moderator Density Coefficient (MDC) were calculated using the GETERA code. The contribution of each isotope presented in the fuel cell was calculated for the temperature range of 20 °C–100 °C at the beginning of the core life. The average values of the FTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were found to be: −2.23E-03, −1.85E-02, −1.96E-02, −1.85E-02 mk/°C respectively. The average values of the MTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −8.91E-03, −1.24E-04, −4.70E-03, 2.10E-03 mk/°C respectively. Finally, the average values of the MDC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −2.06E-01, −2.03E-01, −2.04E-01, −2.03E-01 mk/°C respectively. It's found also that the dominant reactivity coefficient for all types of fuel is the MDC.     

Dry recovery of ceramic uranium dioxide waste material

Three dry recovery routes of ceramic uranium dioxide waste material were studied: (a) calcination of waste material to U₃O₈ and blending with the original UO₂ powder; (b) calcination to U₃O₈, then reduction to UO₂ and blending with the original UO₂ powder; (c) grinding the waste material to a fine powder and blending with the original UO₂ powder. The first route resulted in an increased open porosity and decreased closed porosity of the sintered UO₂ pellets. The second recovery route had almost no effect on the quality of the sintered pellets, while the third route caused an increase in open porosity up to about 15–18 wt% of the added scrap material and increased closed porosity for higher concentrations of the added recovery material. Effects of sieving the recovered scrap powders were also investigated. No effects were observed for the first and second recovery routes, while for the third route more pronounced effects were obtained for coarsely ground powders.     

Heats of formation of (U,Mo)Al3 and U(Al,Si)3

The heats of formation of (U,Mo)Al₃ intermetallic compounds were obtained by measuring the reaction heats of U-Mo/Al dispersion samples by differential scanning calorimetry. Based on literature data for the reaction heats of U₃Si/Al and U₃Si₂/Al dispersion samples, the heats of formation of U(Al,Si)₃ as a function of the Si content were calculated. The heat of formation of (U,Mo)Al₃ becomes less negative as the Mo content increases. Conversely, the heat of formation of U(Al,Si)₃ becomes more negative with increasing Si content.     

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.     

Oxidation of accident tolerant fuel candidates

In this study, the oxidation of various accident tolerant fuel candidates produced under different conditions have been evaluated and compared relative to the reference standard – UO₂. The candidates considered in this study were UN, U₃Si₂, U₃Si₅, and a composite material composed of UN–U₃Si₂. With the spark plasma sintering (SPS) method, it was possible to fabricate samples of UN with varying porosity, as well as a high-density composite of UN–U₃Si₂?(10%). Using thermogravimetry in air, the oxidation behaviors of each material and the various microstructures of UN were assessed. These results reveal that it is possible to fabricate UN to very high densities using the SPS method, such that its resistance to oxidation can be improved compared to U₃Si₅ and UO₂, and compete favorably with the principal ATF candidates, U₃Si₂, which shows a particularly violent reaction under the conditions of this study, and the UN–U₃Si₂?(10%) composite.     

Thermal and mechanical properties of polycrystalline U3Si2 synthesized by spark plasma sintering

U₃Si₂ has been explored as an alternative nuclear fuel material for increased accident tolerance. However, scatter has been reported in the thermal properties possibly because of the pores and impurities within the samples. In the present study, we prepared a polycrystalline U₃Si₂ bulk sample with high density and without impurity, and evaluated its thermal and mechanical properties. The sample was synthesized by arc melting and spark plasma sintering, followed by annealing. The density of the U₃Si₂ pellet was 96% of the theoretical density. The heat capacity was measured and compared with the calculation data. In addition, the measured data were used to evaluate thermal conductivity of U₃Si₂. The measurement data of elastic properties were compared with the theoretical calculation and agreed well. A high thermal conductivity and hardness compare to UO₂make it favorable to anticipated as alternative nuclear fuel.     

Characterization of (Th,U)O2 pellets made by advanced CAP process

Coated Agglomerate Pelletization (CAP) process is being developed by Bhabha Atomic Research Centre (BARC) for the fabrication of ThO₂-UO₂ mixed oxide fuel pellets. In order to improve the microstructures with better microhomogeneity, a study was made to modify the CAP process. The advanced CAP (A-CAP) process is similar to the CAP process except that the co-precipitated powder of mixed oxide, ThO₂-30%UO₂ or ThO₂-50%UO₂, is used for coating instead of U₃O₈ powder. The choice of ThO₂-UO₂ powders as the coating material is advantageous compared to U₃O₈, since the presence of large quantities of ThO₂ in UO₂ powder gives better self-shielding effect. In this paper, ThO₂ containing 4%UO₂ (% in weight) was prepared by the A-CAP process. Property measurements including microstructure and microhomogeneity were made by optical microscopy, scanning electron microscopy (SEM), electron probe microanalysis (EPMA), etc. It was found that the pellets sintered in air at 1400 °C showed a duplex grain structure and those sintered in Ar-8%H₂ at 1650 °C showed a very uniform grain structure with excellent microhomogeneity.     

Investigation of the influence of the initial state of powder,pore-forming additions,and U<Subscript>3</Subscript>O<Subscript>8</Subscript> on the microstructure and strength of fuel pellets

It is shown that the initial state of uranium dioxide powder has no effect on the density, microstructure, and strength of pellets. Pore-forming agents and U₃O₈ used in fabrication lower the pellet strength because their particles are not spherical. To increase pellet strength, it is recommended that U₃O₈ be subjected to special processing to spheroidize the particles before mixing for uranium dioxide powder.     

Irradiation behavior of atomized U–10wt.% Mo alloy aluminum matrix dispersion fuel meat at low temperature

In order to examine the in-reactor behavior of very-high-density dispersion fuels for high flux performance research reactors, U–10wt.% Mo alloy dispersions in an aluminum matrix have been irradiated at low temperature in the Advanced Test Reactor (ATR). The alloy fuel dispersant was produced by a centrifugal atomization process. The fuel shows stable in-reactor irradiation behavior to a fission density of 5×10²⁷ m⁻³ at an irradiation temperature of 65 °C. The fuel–matrix interaction layer growth rate is similar to that observed in uranium-silicide fuels. The fuel particles have a fine and a relatively narrow fission gas bubble size distribution. There appears to be features in the microstrucure that are the result of segregation of the microstructure in to molybdenum rich and depleted regions on solidification.     

XBIC/μ-XRF/μ-XAS analysis of metals precipitation in block-cast solar silicon

The results of the investigations of the interaction between the different impurities in intentionally contaminated block-cast multi-crystalline silicon by means of synchrotron-based microprobe techniques XBIC (X-ray beam induced current), μ-XRF (X-ray fluorescence microscopy) and μ-XAS (X-ray absorption microspectroscopy) recently implemented at beamlines ID-21 and ID-22 of ESRF, Grenoble, are presented. It was found that Si₃N₄/SiC particles frequently observed in the upper part of multi-crystalline Si blocks represent effective sinks for Fe and Cu impurities. The amount of precipitated iron was the same order magnitude both at nitride and carbide particles. The amount of Cu precipitated at the SiC inclusions was significantly larger than that at Si₃N₄ rods. Chemical state of the copper precipitates was identified as copper-rich silicide Cu₃Si. The anneal at 950 °C that is known to enhance oxygen precipitation in silicon was found to accompany with the enhanced formation of nanoscale iron disilicide precipitates both inside the grains and at grain boundaries.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Evolution of microstructure and second-phase particles in Zr–Sn–Nb–Fe alloy tube during Pilger processing

Evolution of microstructure and second-phase particles (SPPs) in Zr–Sn–Nb–Fe alloy tube were investigated during Pilger process using electron backscatter diffraction, secondary electron and transmission electron microscopy imaging techniques. Results show that the Pilger rolled tubes present heterogeneous structures with the C axes of less deformed grains mostly concentrated in the axial direction. During the Pilger rolling, the increase of deformation caused weakening of linear distribution of second-phase particles. The mean diameters of the precipitates are in the range of 70–100 nm in all specimens, and the growth mechanism of SPPs follows second-order kinetics. The grain growth is controlled by Zener pinning in the Pilger rolling–annealing specimens. Clusters containing the Zr(Nb,Fe)₂ and β_Nb precipitates formed in the Zr–1.0Sn–1.0Nb–0.12Fe alloy. Most of the particles located in grain boundaries are the Zr(Nb,Fe)₂ Laves phase with hexagonal structure, and stacking faults have been found in the Zr(Nb,Fe)₂ precipitates. The types, morphology and distribution of precipitates depend on the constituent and structural fluctuations of the nucleation area.     

A new process based agglomeration parameter to characterize ceramic powders

Uranium dioxide powders are made through aqueous chemical route involving precipitation, drying, calcination and reduction. The presence of agglomerates causes powder packing difficulties in the compaction die, and non-uniform and incomplete densification on sintering. To quantify the degree of agglomeration, several authors have proposed ‘Agglomeration Parameters’. The change in BET specific surface area of calcined U₃O₈ upon reduction to UO₂ per unit temperature difference is a simple new measure of agglomeration in uranium dioxide powders.     

Evaluation of precipitates used in strainer head loss testing: Part II. Precipitates by in situ aluminum alloy corrosion

Vertical loop head loss tests were performed with 6061 and 1100 aluminum (Al) alloy plates immersed in borated solution at pH = 9.3 at room temperature and 60 °C. The results suggest that the potential for corrosion of an Al alloy to result in increased head loss across a glass fiber bed may depend on its microstructure, i.e., the size distribution and number density of intermetallic particles that are present in Al matrix and FeSiAl ternary compounds, as well as its Al release rate. Per unit mass of Al removed from solution, the WCAP-16530 aluminum hydroxide (Al(OH)₃) surrogate was more effective in increasing head loss than the Al(OH)₃ precipitates formed in situ by corrosion of Al alloy. However, in choosing a representative amount of surrogate for plant specific testing, consideration should be given to the potential for additional head losses due to intermetallic particles and the apparent reduction in the effective solubility of Al(OH)₃ when intermetallic particles are present.