Uranium Dioxide Pellets Research Articles

Chemical interaction of CsOH vapor with UO2 and Fe-Zr melt

ABSTRACT At TEPCO’s Fukushima Daiichi Nuclear Power Station, it is estimated that considerable amounts of cesium still remain in the reactors from the analysis results using the severe accident analysis codes and the reverse analysis from contaminated water. Since cesium is known to form stable compounds with uranium and zirconium, chemisorption experiments with uranium dioxide pellets and iron-zirconium melts for cesium hydroxide vapor were carried out using the similar method as for stainless steel. As the results, formations of cesium uranate, Cs2UO4, and cesium zirconate, Cs2ZrO3, were confirmed, indicating that cesium was chemisorbed on both of the uranium dioxide pellets and the iron-zirconium melts in an Ar-H2-H2O flow and an Ar-H2 flow, respectively. Therefore, it was considered that cesium released from fuel might be trapped by chemisorption with fuels and/or iron-zirconium melts during light water reactor severe accidents.

Investigation of the influence of the fuel element design parameter on the VVER-1000 reactor axial power peaking factor

The paper presents the results of a numerical study into the efficiency of the fuel element operation in the pressurized water reactor (VVER) core filled with uranium dioxide (UO2) pellets. The investigation results were obtained from a three-dimensional simulation of the fuel element power density. The dependencies of the fuel and fuel cladding temperatures on specific power per cubic meter of fuel are compared. Uranium metal and uranium dioxide have been studied as fuel. Engineering constraints on the safe operation of fuel assemblies have been selected as the determining parameters. The paper analyzes the extent of the radiation heat transfer effects on the fuel element specific power. Equations have been obtained that reflect the dependencies of specific power per cubic meter of fuel on the size of the fuel pellet hole diameter in the maximum heat flux conditions. The COMSOL Multiphysics code, a numerical thermophysical simulation package, was used for the study. Calculations show that an additional uranium-235 enrichment with an increase in the fuel pellet hole diameter at a fixed fuel thermal power leads to a reduced reactor axial temperature field peaking factor.

Diametral compression of short cylinders with a central hole as a method for assessing the tear resistance of brittle materials

The possibility of evaluating the tear resistance of brittle materials by diametral compression test of short cylinders (solid and central with a central hole) has been analyzed. The computational analysis was performed by the finite element method using the ANSYS program. It is shown that the stress distribution in a disk with a hole is similar that without a hole but contains disturbances introduced by a stress concentrator in the form of a hole. The normalized values of the maximum first principal stresses for a disk with a hole exceed the values for a disk without a hole by more than 5 times. The experimental analysis was carried out by testing short cylinders, both solid and with a central hole, made of brittle materials: cast iron and graphite. It is noted that the fracture resistance, determined by the formula recommended by the ASTM D3967 – 95a standard, practically does not differ for solid cast iron samples, and for graphite differs by 1.5 times from the true tear resistance of materials; when testing samples with a central hole, the tear resistance differs from the standard values by a factor of 1.5 and almost 2.5, respectively. The different nature of the sample destruction is also noted: slow controlled rupture of cast iron and dynamic destruction of graphite with the corresponding deformation diagrams. The results of testing fuel pellets of uranium dioxide are given as an example of testing real cylindrical samples with a central hole. It is shown that the test results of ARV-1 graphite samples are in good agreement with the test results of fuel samples. Thus, the possibility of testing small-sized short cylinders according to the diametral compression scheme for indirect assessment of the tensile strength of brittle materials has been confirmed. A calculation formula is proposed for an indirect assessment of the tensile strength of brittle materials based on the results of testing small-sized short cylinders, both with and without a central hole according to the diametral compression scheme.

Direct configuration mode microwave sintering of natural uranium dioxide pellets

The direct microwave sintering of natural uranium dioxide (UO2) pellets at bulk scale is first time successfully demonstrated. The process of the microwave sintering is carried out at 1700 °C, 6h in reducing atmosphere with a sintering cycle spanning over 31h in an indigenously designed and developed sintering set-up deploying 2.45 GHz microwave radiation. The process is demonstrated for obtaining defect-free pellets (devoid of cracks and any plasma induced edge/surface melting or burnouts) in bulk quantity consisting of a batch size of 3 kg (500–550 pellets). The sinter density of pellets was found to be uniform and in the acceptable range of 92–96%TD for nuclear fuel. The comparison of optical microstructures prima facie indicates that the UO2 grain size is smaller in microwave sintered pellets (5–20 μm) as compared to conventionally sintered UO2 pellets (10–40 μm) under identical parameters. The axial and the diameteral shrinkages during densification were observed to be 18–20% which are comparable to the conventional sintering. The microwave sintered UO2 pellets has O/U mole ratio as 2.002. The process is consistent and easily up-scalable to sinter large size of batches so as to meet the industrial scale demands.

Dependence of the thermophysical properties of alloys of the Fe - Cr - Al - Si system on the composition and temperature

Pressurized water reactors (PWR, VVER, EPR, CAP) are currently the backbone of the world's nuclear power industry. The fuel in this type of power plants is uranium dioxide pellets enclosed in a sealed zirconium shell. Loss Of Coolant Accident (LOCA) emergencies can cause partial or complete destruction of zirconium claddings and accumulation of hydrogen in the reactor pressure vessel and reactor building. This is due to the so-called steam-zirconium reaction: an exothermic process of zirconium oxidation in an atmosphere of saturated water vapor at temperatures above 900 °C. The products of this reaction are hydrogen and brittle zirconium oxide, which collapses under the action of its own weight and the weight of the fuel column. Fuel for pressurized water nuclear power reactors that is resistant to a loss of coolant accident is the subject of research around the world. One of the ways of implementation is the use of alloys of the Fe – Cr – Al system as a fuel element cladding material, since they have high corrosion resistance in water. The work is devoted to the determination of the coefficient of linear thermal expansion (CLTE), specific heat capacity, thermal conductivity and the influence of composition on them for Fe – Cr – Al – Si alloys in the temperature range of 200–1000 °C. New data on the dependence of the thermophysical characteristics of alloys of the Fe – Cr – Al – Si system on temperature and chromium concentration have been obtained. It is shown that the CLTE, specific heat capacity and thermal conductivity weakly depend on the chromium content in the studied concentration range, the main contribution of alloying elements is made by the position of the Curie.

Sintering of Industrial Uranium Dioxide Pellets Using Microwave Radiation for Nuclear Fuel Fabrication

In this study, the possibility of sintering industrial pressed uranium dioxide pellets using microwave radiation for the production of nuclear fuel is shown. As a result, the conditions for sintering pellets in an experimental microwave oven (power 2.9 kW, frequency 2.45 GHz) were chosen to ensure that the characteristics of the resulting fuel pellets meet the regulatory requirements for ceramic nuclear fuel, including the following: a density of about 10.44 g/cm3; a volume fraction of open pores of tablets of about 0.1%; an oxygen coefficient of no more than 2.002; hydrogen content of about 0.30 ppm; and the change in density after re-sintering on average no more than 1.16%.

Manufacturing Features and Characteristics of Uranium Dioxide Pellets for Subcritical Assembly Fuel Rods

The influence of technological processes and manufacturing of uranium dioxide fuel pellets for fuel elements for experimental fuel assembly (FA-X) which was designed as an alternative fuel for the nuclear research installation (NRI) "Neutron Source Controlled by Electron Accelerator" were investigated. Unlike standard production processes of UO2 pellets, the special feature fabrication process of this nuclear fuel type is production of uranium dioxide powder with enrichment of 4.4 %wt. of 235U achieved by mixing of two batches of powders with different uranium contents: 0.4 %wt. 235U and 19.7%wt. 235U, as well as ensuring the required tolerance of fuel pellets without the use of machining operations. A set of design and process documentation were developed in the R&D Center at NSC KIPT. Experimental stack of fuel pellets, fuel elements and a pilot fuel assembly FA-X were fabricated and designed to be compatible and interchangeable with VVR-M2 fuel assembly adopted as a standard assembly for the first fuel loading at the "Neutron Source Driven by an Electron Accelerator" FA. As opposition to the variant of VVR-M2 fuel assembly which consisted of three fuel rods of tubular shape with dispersion composition UO2‑Al, FA-X accommodates six fuel rods of pin-type with UO2 pellet which located in the zirconium cladding (E110) as the closest analogue of fuel rods of VVER-1000 power reactor. Inside cladding locate a 500 mm high fuel stack which is secured against displacement by a spacer. In the basic variant of FA-X the fuel pellets are made of UO2 with 235U enrichment near 4.4 %wt.

Proposal for ultrasonic technique for evaluation elastic constants in UO2 pellets

Uranium dioxide pellets (UO2) used as fuel in nuclear power reactors are subjected to high thermal gradients and may undergo swelling during reactor operation. The swelling of UO2 pellets, resulting from the fission products, causes cracking of the material and it is thus critical to study their behaviour. This study evaluated the application of the ultrasonic method in obtaining the elastic constants of UO2 pellets. Due to the difficulties in handling nuclear material, we used tablets of alumina (Al2O3) for a comparative study. For this, two sets of 10 Al2O3pellets with densities of 92% and 96% were fabricated. Ultrasonic transmission technique was used for obtaining measurements of the travel time of longitudinal and transverse ultrasonic waves. Equations relating material density and velocity of the ultrasonic wave to the elasticity, shear and Poisson’s modules allowed the determination of these elastic constants, whose values showed excellent agreement with those available in the literature for Al2O3.

Separate-Effects Tests for Studying Temperature-Gradient-Driven Cracking in UO2 Pellets

A variety of normal operation and accident scenarios can generate thermal stresses large enough to cause cracking in ceramic fuel pellets. Cracking in fuel pellets can lead to reduced heat removal, higher centerline temperatures, and localized stress in the cladding—all which impact fuel performance. It is important to experimentally characterize the thermal and mechanical behaviors in the pellet both before and after cracking, which would help to improve cracking models in fuel performance codes such as BISON. However, in-reactor observation and measurement of cracking are very challenging due to the harsh environment and design of the fuel rods involved. Recently, an experimental pellet-cracking test stand was developed for separate-effects testing of pellet cracking under normal operations and accident temperature conditions using thermal imaging to capture the pellet surface temperatures in order to evaluate the thermal stresses and optical imaging to capture the evolution of cracking in real time. Experiments were performed using depleted uranium dioxide (UO2) pellets, which are useful for collecting valuable data for development and validation of cracking models. A combination of induction and resistance heating was used to create thermal gradients similar to those seen in a reactor environment. Characterization of the pellets was conducted both before and after cracking. The cracking patterns are moderately different from those expected in a typical reactor because of the variations in the thermal conditions and pellet microstructures. However, when the actual conditions of these experiments are reproduced in computational models with sufficient precision, such out-of-pile testing on UO2 pellets provides relevant data for modeling purposes.

Densification of UO2 pellets by microwave sintering using an instrumented multimode cavity

A study of microwave sintering of uranium dioxide pellets has been performed with the aim to densify the material. Microwave sintering heating has been carried out in a 2.45 GHz multimode cavity, designed to fit the different requirements linked to sintering of nuclear ceramics. Hybrid and direct microwave heating (i.e., with and without susceptor) conditions were carried out. In the hybrid condition, thermal cycles at temperatures up to 1700°C, with a heating rate of 20 °C/min and dwell times of 10–60 min could be applied to the samples. The sintered specimens exhibited high values of densities (92.6–94.6% of theoretical density) with a grain size range of 9.1–11.8 μm, which is close to the results obtained with conventional sintering with longer thermal cycles. The pellet tested in the direct configuration only heated until 950°C, showing the impossibility to heat UO2 in a microwave furnace without the help of a susceptor.

Controlled current-rate AC flash sintering of uranium dioxide

Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol−1 and 380 kJ mol−1, respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO2. Both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.

A finite element approach to investigate the thermomechanical behavior of solid and annular type nuclear fuel rods

Reliable prediction of nuclear fuel rod behavior is of prime concern for assessing proper design parameters as well as operating conditions and sustainability aspects in a nuclear reactor. In this study, a numerical analysis has been conducted to investigate the thermomechanical behavior of solid and annular fuel rods individually, considering a pressurized water reactor (PWR). The transient heat conduction equation has been employed for the heat-generating uranium dioxide pellet along with the surrounding zircaloy cladding and solved numerically using the finite element method. Loss of coolant conditions has been simulated by reducing the coolant heat transfer coefficient and compared with the normal operating conditions. In all the cases considered, the annular fuel has shown superior properties in terms of minimized fuel surface heat flux, lower but relatively uniform radial temperature, and thermal strain distribution.

Modeling of oxidation of UO2 pellets

In this study, a mathematical model has been developed for oxidation of pellets of uranium dioxide in the air at the bulk temperature range between 605 and 727 K. Oxidation of this pellet causes an increment in the size of the UO2 particles, decrement of the pellet porosity and breaking-down of this pellet, subsequently. Therefore, the presented model has two stages for this oxidation process, before and after the breaking-down of the pallet. These stages have different physical properties of the pellet and different mechanisms of the oxidation reaction. The predictions of the presented model is validated using the existed experimental data. It is resulted that increasing the bulk temperature increases the rate of the oxidation reaction and decreases the required times for the breaking-down and complete conversion of the pellet. In addition, increasing the size of the pellet increases the required time for breaking-down and decreases the required time complete conversion of the pellet.

Thermal behaviour of caesium implanted in UO2: A comparative study with the xenon behaviour

Xenon and caesium are among the most impacting fission products when studying the nuclear fuel: xenon for its role on the fuel rod thermomechanical behaviour during reactor operation and caesium in the case of atmospheric radioactive release during an accident in a nuclear power plant. This paper focuses on the comparison of caesium and xenon thermal behaviour in polycrystalline uranium dioxide (UO2) pellets. Caesium-133 or xenon-136 stable isotopes were introduced in depleted UO2 samples by ion implantation at a maximum concentration of 0.08 at% at a depth of around 140 nm below the sample surface. Annealing under reducing atmosphere (Ar/H2 5 %) was performed at 1000 °C or 1600 °C, which corresponds respectively to a representative temperature during nuclear reactor operation (at the centre of the fuel pellets) and during an accident. The caesium migration in UO2 was investigated by Secondary Ion Mass Spectrometry and compared to the thermal behaviour of xenon in UO2 at 1600 °C. Transmission Electron Microscopy was performed in order to characterise UO2 microstructure before and after annealing. The results indicate that caesium has a different behaviour than xenon with which it is often compared for its release from the nuclear fuel. In particular, we highlight a difference between the growth kinetics of caesium and of xenon bubbles at 1600 °C which can be correlated to the availability of thermal vacancies in UO2 and to the different ability of Xe and Cs atoms for thermal resolution.

Bundle geometry design influence on fuel burnup and performances in a CANDU reactor

The paper evaluates burnup and fuel performance aspects of some advanced CANada Deuterium Uranium (CANDU11CANDU=CANada Deuterium Uranium; 2CFS=Core Follow Simulations; 3ICB=Instantaneous Critical Burnup; 4TA=Time Average; 5ADB=Average Discharge Burnup; 6ZCU=Zone Control Units; 7FPD=Full Power Day; 8RAB=Refuelling Average Burnup; 9(M)OELP(B)=(Maximum) Outer Element Linear Power (Boost)) bundle geometry designs with more than 37 fuel elements and containing natural uranium dioxide pellets. Three advanced bundle geometries have been considered: the 43, 52 and 61-rods ones. Lattice calculations were performed to generate cross section tables with respect to the burnup and to find out instantaneous critical burnups. The burnup efficiency expressed as Uranium consumption for a reference period was evaluated through Core Follow Simulations (CFS). Outer element linear powers and power boosts together with thermo-mechanical parameters were calculated and discussed. The conclusions underline the burnup efficiency and associated uncertainties along with normal operation safety improvements brought by the advanced CANDU bundle geometry designs with increased numbers of fuel elements. Proposals for future work are made in order to get more reliable results using advanced modelling computer tools.

Evaluation of the Possibility of Fabricating Uranium-Molybdenum Fuel for VVER by Powder Metallurgy Methods

It is shown that the technological scheme for fabricating uranium dioxide pellets is suitable for fabricating uranium-molybdenum pellets. It permits using the standard powder metallurgy methods to fabricate pellets from uranium-molybdenum powder obtained by thermocentrifugal atomization of ingots of uranium-molybdenum alloy. The proposed technology uses standard equipment operating in easily reproducible regimes.

Influence of strain conditions on the grain sub-structuration in crept uranium dioxide pellets

Stoichiometric uranium dioxide (UO2) pellets were deformed by uniaxial compression creep tests at 1500 °C. Strains comprised between 3% and 11% were applied at strain rates ranging from ∼ 5 × 10−6 s−1 to 70 × 10−6 s−1 to deform pellets in the dislocational creep regime. An optimized protocol for Electron BackScattered Diffraction (EBSD) data acquisition and processing was applied to detect and quantify low and very low angle sub-boundaries disoriented down to 0.25° inside the prior UO2 grains. Three EBSD based parameters were calculated and are efficient to quantify the evolution of the deformed microstructure with the deformation conditions: the linear fraction of the sub-boundaries, the corresponding Geometrically Necessary Dislocations (GNDs) density and the resulting mean sub-grain size. The linear fraction of sub-boundaries (and their transcription in GNDs) increased significantly with the strain level and rate. New very low angle boundaries disoriented by less than 1° were continuously created whereas already existing sub-boundaries increased continuously their disorientation. This microstructural evolution can be described as a dynamic recovery mechanism by progressive increase of sub-boundaries disorientation angle.

Production of more ideal uranium trioxide microspheres for the sol-gel microsphere pelletization process without the use of carbon

Renewed interest by the U.S. Department of Energy in the sol-gel microsphere pelletization process led to an evaluation of dried uranium trioxide (UO3) microspheres produced without the use of carbon by the internal gelation process at Oak Ridge National Laboratory. An earlier study has shown that less dense UO3 microspheres led to uranium dioxide pellets with higher densities. This effort identified feed formulations that produced dried UO3 microspheres with the lowest tap densities for gelation temperatures of 333, 343, 353, and 363 K. The tap densities of these dried UO3 microspheres ranged from 680 to 820 kg/m3, which were significantly lower than the previously reported low UO3 density of 900 kg/m3. This study confirmed an earlier observation that the effects of gelation conditions on the properties of the gel sphere are complex since only two of the four best formulations were similar in composition. The need to age the wet gel spheres in the gelation medium for 20 min at the end of the experiment was also confirmed.

Permeability of observed three dimensional fracture networks in spent fuel pins

The three-dimensional fracture network within a spent fuel pin is characterised using sequential grinding, and its permeability is numerically estimated. Advanced Gas-cooled Reactors (AGRs) produce spent fuel pins consisting of an outer steel cladding enclosing ceramic uranium dioxide (UO2) pellets. During irradiation, fuel pellets may undergo fracturing due to thermal, densification and swelling effects. Fracture patterns are usually observed on the surface of the pellet or through a cross section or longitudinal plane along the pellet. In this work, the three-dimensional fracture pattern within the pellet is characterised using an optical microscope. The pellet is progressively ground and polished, providing sequential cross sections, which together yield a three-dimensional discrete fracture pattern. Multiple large fractures grow to connect the cladding to the internal region of the pellet. Multiple surface fractures are observed that do not penetrate into the matrix of the pellet. The porosity of the UO2 and apertures of the fractures are estimated by sampling microscopic images. Darcy flow is numerically solved using the finite element method, computing flow through the matrix and fractures simultaneously. The equivalent tensorial permeability of the system is estimated for various approximate fracture apertures. The fracture network raises the permeability of the pellet by an order of magnitude.

Synthesis of High-Density Pellets of Uranium Dioxide by Spark Plasma Sintering in Dies of Different Types

A procedure for electric pulse consolidation of commercial UO2 powders of various ceramic grades into high-density fuel pellets using dies of various types was studied. The influence of the parameters of spark plasma sintering (SPS) on the chemical composition of the initial powders and quality of ready UO2 pellets was examined. The main advantages and drawbacks of the SPS process for consolidation of UO2 powder in standard graphite dies were revealed. A novel alternative procedure for SPS of UO2-based fuel pellets in Mo-based dies was suggested. High quality of the pellets (density 97.5–98.4% of theoretical, no C impurity, mean grain size no more than 3 μm) and mild sintering conditions (1100°С, 141.5 MPa, 25 min) make the process highly promising. The results obtained are of interest for the development and possible implementation of an industrial SPS process for production of ceramic nuclear fuel.