Specification of limiting sizes of PuO2-microheterogeneities in pellets of mixed oxide fuels

A general study was performed of local overheating in mechanically-blended uranium and plutonium mixed oxide fuels on the internal surface of the cladding due to the presence of adjacent PuO₂ particles or PuO₂-rich zones in steady-state conditions. Several idealized geometries of the particles have been considered, generalizing the results of a previous paper. Using a two-dimensional finite element heat conduction code, the steady-state spatial distribution of the dimensionless increments of temperature on the internal surface of the cladding was obtained in terms of the relevant dimensionless parameters of the problem. The information obtained allows the designer of this kind of mixed oxide fuels to solve various thermal problems concerned with quality control.     

Radiation chemical synthesis and characterization of UO2 nanoparticles

In a deep repository for spent nuclear fuel, U(VI)(aq) released upon dissolution of the fuel matrix could, in reducing parts of the system, be converted to U(IV) species which might coalesce and form nanometer-sized UO₂ particles. This type of particles is expected to have different properties compared to bulk UO₂(s). Hence, their properties, in particular the capacity for oxidant consumption, must be investigated in order to assess the effects of formation of such particles in a deep repository. In this work, methods for radiation chemical synthesis of nanometer-sized UO₂ particles, by electron- and γ-irradiation of U(VI) solutions, are presented. Electron-irradiation proved to be the most efficient method, showing high conversions of U(VI) and yielding small particles with a narrow size distribution (22-35 nm). Stable colloidal suspensions were obtained at low pH and ionic strength (pH 3, I = 0.03). Furthermore, the reactivity of the produced UO₂ particles towards H₂O₂ is investigated. The U(IV) fraction in the produced particles was found to be ∼20% of the total uranium content, and the results show that the UO₂ nanoparticles are significantly more reactive than micrometer-sized UO₂ when it comes to H₂O₂ consumption, the major part of the H₂O₂ being catalytically decomposed on the particle surface.     

Study of homogenisation and cation interdiffusion in mixed UO2-PuO2 compacts by X-ray diffraction

Homogenisation in mixed UO₂-PuO₂ compacts has been studied by X-ray diffraction. It is observed that the homogenisation proceeds, mainly, by the assimilation of UO₂ into PuO₂. This near one-way flow of material, from UO₂ to PuO₂, is shown to be due to high activity (large BET surface area) of the PuO₂ powder as compared with that of the UO₂ powder.An X-ray line profile analysis method of determining various mixed composition fractions in sintered mixed compacts has been used to evaluate homogeneity in terms of the fraction of UO₂ that has gone into PuO₂. A concentric core-shell diffusion model, in which UO₂ forms a solute core and PuO₂ forms a solvent shell, was used to determine cation interdiffusion coefficients from the homogenisation data. The temperature dependence of cation diffusivity in the range 1573–1873 K is obtained as $\text{D = 2.55 × 10}{}^{\mathrm{?11}}\text{exp(?2.22 × 10}{}^5\text{/8.31 T)}$ m²/s. The low value (222 kJ/mole) of activation energy for cation interdiffusion is attributed to the hypostoichiometry of the mixed compacts studied.The diffusivity values at 1573 and 1673 K separately give an activation energy of 126 kJ/mole, which suggests grain-boundary diffusion as the primary mechanism of homogenisation in this temperature range.     

Irradiation-induced creep of mechanically blended porous UO2-PuO2 fuel

With regard to the behaviour of fast breeder reactor fuel, irradiation creep of mechanically blended, porous U_natO₂-15% PuO₂ was investigated. Some results for UO₂ are also quoted to clarify the dependence of creep rate on stress and temperature. The sintered density of the UO₂-PuO₂ samples amounted to 86% TD and 93,5% TD, their irradiation temperatures were between 300 and 1000°C, the stress in the samples at 15 and 40 MN/m², the fission rates between 2.5 and $\text{5 × 10}{}^{\mathrm{?9}}$ f/(U + Pu)-atom · s, and the maximum burnup at about 1%. The creep rates of UO₂-PuO₂ are much higher than previously measured on UO₂ of high density, but there was a good correspondence of the stress and temperature dependence. The difference of the creep rates cannot be explained only by the porosity of the UO₂-PuO₂ samples. Therefore the PuO₂ portion of the fuel, whose distribution is heavily inhomogeneous, is treated as additional “effective” porosity. By this means a suitable interpretation is obtained for the results below about 650°C. At higher temperatures, UO₂-PuO₂ of 86% TD showed a rapid initial densification up to about 93% TD, apparently together with a simultaneous homogenization of the fission-density distribution. The results measured could be interpreted without considering an influence of the Pu-content as such.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO₂-yttria stabilized zirconia (YSZ: (Zr,Y)O_2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as ²³⁸U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO₂ HTGR case by adding Er in the fuel. Burnup calculations on PuO_2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for ²³⁹Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO₂ HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.     

A fabrication technique for a UO2 pellet consisting of UO2 grains and a continuous W channel on the grain boundary

A new fabrication process of UO₂-W composite fuel has been studied in order to improve the thermal conductivity of the UO₂ pellet by the addition of a small amount of W. A fabrication process was designed from the phase equilibria among tungsten, tungsten oxides and UO₂. The conventionally sintered UO₂ pellet which contains W particles is heat-treated in an oxidizing gas and then in a reducing gas. In the oxidizing heat-treatment W particles are oxidized and liquid tungsten oxide penetrates within the UO₂ grain boundary, and in the reducing heat-treatment liquid oxide is transformed to solid tungsten which forms a continuous channel along the UO₂ grain boundary. This developed technique can provide a continuous W channel covering UO₂ grains for a UO₂-W composite fuel even with a small amount of a metal phase - below 6 vol.%. The thermal diffusivity of the UO₂-6 vol.%W cermet composite increases by about 80% when compared with that of a pure UO₂ pellet.     

A theoretical study of electronic and valence properties of U/PuO2

We consider the basic electronic and thermodynamic properties of the U/PuO₂ phase, paying special attention to the contrasts between the behaviour of this material and that of UO₂. We also report calculated ionisation potentials for Pu which play an important rôle in our analysis.     

The effects of alpha-radiolysis on UO2 dissolution determined from batch experiments with Pu-doped UO2

The effects of alpha dose-rate on UO₂ dissolution were investigated by performing dissolution experiments with ²³⁸Pu-doped UO₂ materials containing nominal alpha-activity levels of ∼1-100 Ci/kg UO₂ (actual levels 0.4-80 Ci/kg UO₂), in 0.1 M NaClO₄ and in 0.1 M NaClO₄ + 0.1 M carbonate. Dissolution rates increased less than 10-fold for an almost 100-fold increase in doping level and fall within the range of predictions of the Mixed Potential Model (a detailed mechanistic model for used fuel dissolution). Dissolution rates were lower in carbonate-free solutions and enrichment of ²³⁸Pu on the UO₂ surface was suggested in carbonate solutions. Effective G values, defined as the ratio of the total amount of U dissolved divided by the maximum possible amount of U dissolved by radiolytically produced H₂O₂, increased with decreasing doping levels. This suggests that the dissolution reaction at high dose rates is limited by the reaction rate between UO₂ and H₂O₂, but becomes increasingly limited by the rate of production of H₂O₂ at lower dose rates.     

Raman spectroscopy characterization of actinide oxides (U1−yPuy)O2: Resistance to oxidation by the laser beam and examination of defects

Structural changes in four (U_1−yPu_y)O₂ materials with very different plutonium concentrations (0 ? y ? 1) and damage levels (up to 110 dpa) were studied by Raman spectroscopy. The novel experimental approach developed for this purpose consisted in using a laser beam as a heat source to assess the reactivity and structural changes of these materials according to the power supplied locally by the laser. The experiments were carried out in air and in water with or without hydrogen peroxide. As expected, the material response to oxidation in air depends on the plutonium content of the test oxide. At the highest power levels U₃O₈ generally forms with UO₂ whereas no significant change in the spectra indicating oxidation is observed for samples with high plutonium content (²³⁹PuO₂). Samples containing 25 wt.% plutonium exhibit intermediate behavior, typified mainly by a higher-intensity 632 cm⁻¹ peak and the disappearance of the 1LO peak at 575 cm⁻¹. This can be attributed to the presence of anion sublattice defects without any formation of higher oxides. The range of materials examined also allowed us to distinguish partly the chemical effects of alpha self-irradiation. The results obtained with water and hydrogen peroxide (a water radiolysis product) on a severely damaged ²³⁸PuO₂ specimen highlight a specific behavior, observed for the first time.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

Theoretical fuel element modelling at CNEA

Computer codes and theoretical developments aiming to model nuclear reactor fuel elements at CNEA are reviewed. The codes PELT, VAINA and BACO for the overall fuel behaviour, as well as the finite element systems ELASTEF, PLASTEF and CTR are described. The influence on the BACO code predictions of including a fuel cracking model is discussed. Also, some examples of the calculated fuel cladding contact pressure are shown for different situations. Applications of the finite element systems to calculate the stress concentration at the skids of the Central Nuclear Atucha fuel rods and to predict local thermal effects of PuO₂ particles in a UO₂ fuel are discussed.     

H2 inhibition of radiation induced dissolution of spent nuclear fuel

In order to elucidate the effect of noble metal clusters in spent nuclear fuel on the kinetics of radiation induced spent fuel dissolution we have used Pd particle doped UO₂ pellets. The catalytic effect of Pd particles on the kinetics of radiation induced dissolution of UO₂ during γ-irradiation in containing solutions purged with N₂ and H₂ was studied in this work. Four pellets with Pd concentrations of 0%, 0.1%, 1% and 3% were produced to mimic spent nuclear fuel. The pellets were placed in 10 mM aqueous solutions and γ-irradiated, and the dissolution of was measured spectrophotometrically as a function of time. Under N₂ atmosphere, 3% Pd prevent the dissolution of uranium by reduction with the radiolytically produced H₂, while the other pellets show a rate of dissolution of around 1.6 × 10⁻⁹ mol m⁻² s⁻¹. Under H₂ atmosphere already 0.1% Pd effectively prevents the dissolution of uranium, while the rate of dissolution for the pellet without Pd is 1.4 × 10⁻⁹ mol m⁻² s⁻¹. It is also shown in experiments without radiation in aqueous solutions containing H₂O₂ and O₂ that ?-particles catalyze the oxidation of the UO₂ matrix by these molecular oxidants, and that the kinetics of the catalyzed reactions is close to diffusion controlled.     

Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

Plutonium dioxide (PuO₂) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO₂ at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO₂ were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO₂ were investigated up to 1500 K.     

Optical properties of UO2 and PuO2

We perform first-principles calculations of electronic structure and optical properties for UO₂ and PuO₂ based on the density functional theory using the generalized gradient approximation (GGA) + U scheme. The main features in orbital-resolved partial density of states for occupied f and p orbitals, unoccupied d orbitals, and related gaps are well reproduced compared to experimental observations. Based on the satisfactory ground-state electronic structure calculations, the dynamical dielectric function and related optical spectra, i.e., the reflectivity, adsorption coefficient, energy-loss, and refractive index spectrum, are obtained. These results are consistent with the available experiments.     

Thermal-hydraulic and safety aspects of UO2 fuel for low-power reactors

The low enriched uranium UO₂ (about 19.75% U²³⁵) fuel is proposed to be used in low-power research reactors. The thermal-hydraulic and dynamic characteristics are examined in this paper. The fuel behaves similarly to the actual highly enriched uranium fuel in the normal daily operation for both Miniature Neutron Source Reactors and SLOWPOKEs, the cladding temperature reaching about 60 °C. During the simulation of a design basis accident the reactor power peak and temperatures are found to be higher than in the case of the highly enriched uranium fuel for MNSRs, the power peak touching 135 kW, and the cladding temperature reaching over 110 °C in this case. Nevertheless the fuel can be safely used in these reactors.     

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

The codes devised and used in India for the design of fuel for their Pressurized Heavy Water Reactor (PHWR) programme are described. The scheme includes the use of collapsible fuel cladding for improved neutron economy.This code is made with reference to collapsible clad UO₂ fuel elements. This evaluates sheath strain and fission gas pressure. The fuel expansion is calculated by a two zone model which assumes that above a certain temperature the UO₂ deforms plastically and below that temperature it cracks radially and behaves as an elastic solid; the plastic core is under compression. The pellet clad gap conductance is calculated by using a modified Ross and Stoute model considering the effects of fuel and clad thermal expansion, fission gas release, dilution of filler gas and irradiation swelling. Stress relaxation of the sheath and its effect on fuel sheath contact pressure is also considered for arriving at the end result.     

Prospective Power-Reactor Fuel Cycles Based on Water-Free Reprocessing of Spent Fuel

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO₂, PuO₂ or (U, Pu)O₂ from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O₂ fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.     

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.