Uranium Silicide Research Articles

Surface and interface study of U/Si (1 1 1)

Thin uranium films were deposited on Si (111)—7×7 surface by e-beam evaporation. The surface and interface of U/Si (111) with different annealing temperatures were studied by scanning tunneling microscopy (STM), low-energy electron diffraction (LEED), reflection high energy electron diffraction (RHEED) and photoemission spectroscopy. The formation of 2D uranium silicide (USi1.67) film was confirmed by the presence of a sharp 1×1 LEED pattern, and the surface morphology of this phase displays triangular layered structures. A new superstructure was found for the U–Si system when annealing the interface at 1000K. Further annealing of the interface leads to the appearance of large area of Si (111)—7×7 reconstruction surface with high islands on it.

First-principles investigations of the physical properties of binary uranium silicide alloys

The structural, elastic properties and the Debye temperature of binary Uranium Silicide (U-Si) alloys are investigated by using the first-principles plane-wave pseudopotential density function theory within the generalized gradient approximation (GGA). The ground states properties are found to agree with the available experimental data. The mechanical properties like shear modulus, Young’s modulus, Poisson’s ratio σ and ratio B/G are also calculated. Finally, The averaged sound velocity (vm), the longitudinal sound velocity (vl), transverse sound velocity (vt) and the Debye temperature (θD) are obtained. However, the theoretical values are slightly different from few existed experiment data because the latter was obtained at room temperature while the former one at 0K.

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58gU/cm3) and enrichment (93%) of HEU fuel (U–Al), which may result in the changes of heat transfer response and neutronic characteristic in the core. So it is necessary to objectively re-assess the feasibility of LEU silicide fuel core in KUR by using various numerical simulation codes. This paper established a detailed simulation model for the LEU silicide core and provided the safety analyses for the reactivity insertion transients in the core by using EUREKA-2/RR code. Although the EUREKA-2/RR code is a proven and trusted code, its validity was further confirmed by the comparison with the predictions from another two thermal hydraulic codes, COOLOD-N2 and THYDE-W at steady state operation. The steady state simulation also verified the feasibility of KUR to be operated at rated thermal power of 5MW. In view of the core loading patterns, the operational conditions and characteristics of the reactor protection system in KUR, the accidental control rod withdrawal transients at natural circulation and forced circulation modes, the cold water injection induced reactivity insertion transient and the reactivity insertion transient due to removal of irradiation samples were conservatively analyzed and their transient characteristic parameters such as core power, fuel temperature, cladding temperature, primary coolant temperature and departure from nucleate boiling ratio (DNBR) due to the different ways and magnitudes of reactivity insertions were focused in this study. The analytical results indicate that the quick power excursions initiated by the reactivity insertion can be safely suppressed by the reactor protection system of KUR in various initial power levels and different operational modes (natural circulation and forced circulation modes). No boiling and no burnout on fuel cladding surface and no blister in the fuel meat happens and KUR is safe in all of these reactivity insertion transients if the reactor protection system of KUR works in its minimum degree.

A study on improving the performance of a research reactor's equilibrium core

Utilizing low enriched uranium silicide fuel (U3Si2-Al) of existing uranium density (3.285 g/cm3), different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of ?unit flux time cycle length per 235U mass per cycle?. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3) without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

First principles investigations of the electronic structure and chemical bonding of U3Si2C2 – A uranium silicide–carbide with the rare [SiC] unit

The electronic structure of U3Si2C2, with the rare [SiC] unit is examined from ab initio with an assessment of the properties of chemical bonding. We show that plain GGA fails describing the experimental lattice parameters and the electronic structure. A better agreement with experiment (crystal determination and magnetic properties) is obtained with the GGA+U method and U=4eV. The energy–volume equation of state and the set of elastic constants are obtained showing incompressibility along the c-axis with U–C–Si alignment and a brittle material. Bonding of U1 and U2 selectively with Si and C and Si–C bonds are remarkable

High pressure studies on uranium and thorium silicide compounds: Experiment and theory

The actinide silicides ThSi, USi and USi2 have been studied under high pressure using both theory and experiment. High pressure synchrotron X-ray diffraction experiments were performed on polycrystalline samples in diamond anvil cells at room temperature and for pressures up to 54, 52 and 26GPa, for ThSi, USi and USi2, respectively. At ambient conditions, the uranium silicides crystallize in tetragonal structures (space groups: I4/mmm for USi and I41/amd for USi2), while ThSi adopts an orthorhombic structure (space group: Pbnm) (including an anharmonic analysis of the silicon). These structures are found to be stable with no structural transitions observed up to the highest pressures achieved. The zero-pressure bulk modulus B0 and its pressure derivative B0′ at ambient pressure are obtained from the measured P–V relations. The experiments are accompanied by first principles calculations using the full-potential linear muffin-tin orbital method within the generalized gradient approximation for exchange–correlation effects. Experimental results are well reproduced by the calculated equation of state and ground state properties.

Thermal evaluation of uranium silicide miniplates irradiated at high heat flux

The Gas Test Loop (GTL)-1 irradiation experiment was conducted in the Advanced Test Reactor (ATR) to assess corrosion performance of proposed booster fuel at heat flux levels ∼30% above the design operating condition. Sixteen miniplates fabricated from 25% enriched, high-density (4.8gU/cm3) U3Si2/Al dispersion fuel with 6061 aluminum cladding were subjected to peak beginning of cycle (BOC) heat fluxes ranging from 411 to 593W/cm2. No adverse impacts to the miniplates were observed at these high heat flux levels. A detailed finite element model was constructed to calculate temperatures and heat flux for an as-run cycle average effective ATR south lobe power of 25.4MW(t). Miniplate heat flux levels and fuel, cladding, hydroxide, and coolant–hydroxide interface temperatures were calculated using the average hydroxide thickness on each miniplate measured during post-irradiation examination. The purpose of this study was to obtain a best estimate of the as-run experiment temperatures to aid in establishing acceptable heat flux levels and designing fuel qualification experiments for this fuel type.

Uncertainty measurement evaluation of WDXRF and EDXRF techniques for the Si and Utotal determination in U3Si2 nuclear fuel

Uranium silicide (U3Si2), 20% 235U enriched powder, is an intermetallic compound used as nuclear fuel material, which is the state-of-the-art among nuclear fuel materials used in modern research reactors. It is produced by IPEN and used as nuclear fuel of the IEA-R1 reactor (IPEN/CNEN, Sao Paulo, Brazil); U3Si2 has 92.3 wt% Utotal and 7.7 wt% Si. The qualification of this material requires chemical and physical tests such as Si and Utotal content, isotope ratio, impurities, density, specific surface area and particle size determination. The Si and Utotal were determined by gravimetric and volumetric procedures. Usually, these classical methods require a long time for analyses and are expensive. The objective of this study was to establish a fast and efficient analytical method to meet ISO/IEC 17025:2005 requirements in the Si and Utotal determination. The X-ray fluorescence techniques (XRF) were chosen to allow a direct and non-destructive analysis, what is the main advantage compared to other instrumental techniques, since previous chemical treatments are not necessary. In this study, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X-ray fluorescence techniques was evaluated. Furthermore, two different sample preparation procedures, plain powdered and pressed powdered, were evaluated. Statistical tools were used to evaluate the results and a comparison between these results and the conventional methods was done.

ChemInform Abstract: Characterization of Novel Ternary Uranium Silicides and Germanides with the U2Mo3Si4 Structure Type in the U-(Mo, W, V)-(Si, Ge) Systems.

ChemInform is a weekly Abstracting Service, delivering concise information at a glance that was extracted from about 100 leading journals. To access a ChemInform Abstract of an article which was published elsewhere, please select a “Full Text” option. The original article is trackable via the “References” option.

Development of powder metallurgy technique for synthesis of U 3Si 2 dispersoid

The Indian research reactor APSARA, which uses HEU based UAl 3 fuel dispersed in aluminium matrix, is to undergo core conversion to LEU based U 3Si 2 dispersoid in aluminium matrix fuel. Work has been initiated at BARC, to prepare uranium silicide (U 3Si 2) compound using powder processing route in place of vacuum induction melting or arc melting route. This paper describes the preparation flow-sheet of the U 3Si 2 dispersoid by an innovative powder processing route with uranium metal powder and silicon powder as starting materials.

Fabrication of U<sub>3</sub>Si<sub>2</sub> Powder for Fuels Used in IEA-R1 Nuclear Research Reactor

Powdered uranium silicide (U3Si2) 20% U235 enriched is an intermetallic compound used as nuclear fuel material dispersed in aluminum to be the meat of fuel elements. U3Si2 powder is the state-of-the-art as nuclear fuel material mostly used in modern research reactors. Its recent established fabrication in IPEN replaced the previous ceramic powder U3O8 used in the fuel of IEAR1 (IPEN/CNEN, São Paulo, Brazil). The U3Si2 is a compound with 92.3%wtU and 7.7%wtSi. Its production is made by induction furnace melting using metallic uranium, produced by magnesiothermic reaction, and pure silicon. The induction furnace melts under argon controlled environment using zirconia crucible. Homogenization of liquid bath at 1800°C is a compromise between crucible resistance and homogenized melting, avoiding hazardous happenings. IPEN produced its first lot of enriched U3Si2 in September 2004, with a continuous fabrication ever since. This research work represents the ability of having fully Brazilian supply of this strategic and high cost nuclear material. The fuel quality meets the world quality standards required by International Atomic Energy Agency (IAEA) and RERTR standards. Brazilian production of U3Si2 powder not only closed the fuel cycle, from uranium mineral to fuel element, but also allowed higher productivity of nuclear medicine radioisotopes by IEA-R1.

Development of high density uranium compounds and alloys as dispersion fuel for research and test reactors

Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) programme, which was launched in the late 70’s to reduce the threat of nuclear proliferation. Various compounds of uranium were considered and studied for the purpose but the best fuel compound identified for the development of low enriched uranium base fuel was uranium silicide (U3Si2). The fabrication method qualified for making U3Si2-Al dispersion fuel was giving a maximum heavy atom density in the range of 4.8 g/cm3 to 6.0 g/cm3. Under the RERTR programme an ambitious goal of achieving heavy metal density of 8.0 to 9.0 g/cm3 in aluminium based dispersion fuel was set and to realize this high density γ - stabilized uranium alloy (e.g. U-Mo, U-Zr etc.) powder in conjunction with advanced fabrication techniques.

Transient behaviour of low-enriched uranium silicide plate-type fuel for research reactors during a reactivity-initiated accident

In many research reactors and test reactors, a low enriched (<20 wt% 235U) silicide fuel is to be used in the replacement of a high (93wt% 235U) to medium (45wt% 235U) enrichment aluminide fuel bec...

Irradiation tests for U 3Si–Al dispersion fuels with aluminum cladding

The HANARO fuel element is made of a cylindrical fuel meat, top and bottom aluminum end plugs and an aluminum cladding with eight longitudinal fins. The fuel meat of each fuel element consists of a dispersion of small particles of a high density uranium silicide (U 3Si) compound in a continuous aluminum matrix. To verify the irradiation performance of the HANARO fuel at a high power and a high burnup, in-pile irradiation tests were performed in the HANARO core. Detailed non-destructive and destructive post-irradiation examinations were conducted. It was verified through the irradiation tests that the HANARO fuel maintains a proper in-pile performance and integrity even at a high power of 121 kW/m and up to a high burnup of 85 at.% U-235.

A study of reduction in fuel requirement and enhancement of neutron flux in typical MTR type research reactors

Pakistan Research Reactor-1 (PARR-1) is a typical swimming pool type MTR utilizing low enriched uranium (19.99% in 235U) silicide dispersion fuel of density 3.28 gU/cm 3. The benefits of loading available and upcoming higher density fuels in the existing PARR-1 core have been explored in this study with constraints of no change in the existing moderator/coolant channel width and existing reactor systems. The study was conducted by employing the standard reactor physics simulation codes WIMS-D/4, CITATION, and burnup analyses code FCAP along with a reactor thermal hydraulics simulation code PARET. The study reveals that by directly replacing the fuel currently in use in PARR-1 core with a similar type of fuel of density 4.8 gU/cm 3, an equilibrium core slightly smaller than the existing equilibrium core of PARR-1 can be established. The new equilibrium core can provide neutron fluxes similar to those that are available in the existing equilibrium core of PARR-1 but will require about 14.4 kg less LEU fuel for its one effective full power year operation at 10 MW. Size of this new equilibrium core can be reduced to achieve about 47% higher thermal neutron fluxes at the irradiation sites without increasing the existing cost of producing neutron fluxes. In case of new cores, fuel cycle length is also six effective full power days (EFPDs) larger than that of the existing equilibrium core. Thermal hydraulic analysis results show that these suggested cores can be operated safely at 10 MW with the existing coolant flow rate of 1000 m 3/h. The study was further extended by considering direct substitution of higher density fuels, being developed under the RERTR program, in the proposed core. This extended study revealed that the utilization of further higher density fuels being developed under the RERTR program is not feasible from the viewpoint of relative production cost of neutron flux.

Irradiation behaviour of uranium silicide compounds

A study of the irradiation behaviour of uranium silicide and other related inter-metallic uranium compounds is presented. This study was motivated by the recent discovery that U 3Si 2 undergoes a crystalline to amorphous transformation during irradiation. Such information renders a previously developed fuel swelling model based on the crystalline state of U 3Si 2 invalid. This is of particular significance since low enriched U 3Si 2 dispersion fuels are widely used in research reactors. While such a finding does not alter the well established, stable and benign behaviour of U 3Si 2 during irradiation, it does indicate that a different interpretation of that behaviour is required.

A model for fission-gas-bubble behavior in amorphous uranium silicide compounds

A model for the behavior of fission gas in irradiated amorphous materials is developed. The model proposes that gas bubble nucleation occurs within shear bands initiated around free volume regions. Small gas–atom clusters that form within these regions are susceptible to dissolution by forces generated by the plastic flow of material around the cluster. The bubble coarsening process depends on the materials viscosity and on irradiation-induced re-solution. The bubble distribution eventually reaches a point where larger bubbles from the tail of the evolving lognormal size distribution begin to contact the more numerous nanometer-sized bubbles from the peak region. This condition defines the knee in the swelling curve. The fission density at which the knee occurs is a function of fission rate. Calculations for the behavior of the knee, swelling, and the fraction of gas in bubbles in irradiated U 3Si 2 intermetallic compounds are compared to measured quantities.

Photoemission study of the U/Si(111) interface

The electronic and geometric structures of thin uranium overlayers deposited on a clean Si(111)-(7×7) surface have been studied by X-ray photoelectron spectroscopy (XPS) and low-energy electron diffraction (LEED). Three different stages of interaction between the deposited uranium and the silicon substrate are found, depending on the uranium coverage. A transition from localized to delocalized U5f states is observed as the uranium thickness increases in the room-temperature depositions. Annealing of the substrate at 700°C results in clear spots in the LEED observations, suggesting the possible formation of ordered uranium silicide phases at the interface.

DART Model for Irradiation-Induced Swelling of Uranium Silicide Dispersion Fuel Elements

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U3SiAl-Al and U3Si2-Al in variously designed dispersion fuel elements.