High resolution TEM of high burnup UO2 fuel

High-resolution TEM (HRTEM) observations and nano-area EDX analyses were carried out on small intragranular bubbles in the outer region of high burnup UO₂ pellets. Sample was prepered from the outer region of UO₂ pellet, which had been irradiated to the pellet average burnups of 49 GWd/t in a BWR. HRTEM observations and element analyses were made with a 200 KV cold-type field emission TEM (Hitachi FE-2000) having an ultra-thin window EDX (Noran Voyager). Lattice image and nano-area EDX results indicate the presence of 4-8 nm size Xe-Kr bubbles along with fission products of five metal particles, Mo-Tc-Ru-Rh-Pd. Nano-diffraction patterns from bubbles show two different new patterns besides matrix UO₂. From the Xe/U proportion obtained by nano-area EDX peak and nano-diffraction patterns from bubbles, it was concluded that Xe in the small bubbles was present in a solid or near solid state at very high pressure. Furthermore, from the results of high resolution images and diffractions obtained from recrystallized grains in rim structure region, neighboring recrystallized grains were clarified to be present with high angle grain boundaries.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.     

The creep of UO2 fuel doped with Nb2O5

The creep of UO₂ containing small additions of Nb₂O₅ has been investigated in the stress range 0.5–90 MN/m² at temperatures between 1422 and 1573 K. The functional dependence of the creep rate of five dopant concentrations up to 0.8 mol% Nb₂O₅ has been examined and it was established that in all the materials the secondary creep rate could be represented by the equation /.εkT = Aσⁿexp(?Q/RT), where /.ε is the steady state creep rate per hour, Q the activation energy and A and n are constants for each material. It was observed that Nb₂O₅ additions can cause a dramatic increase in the steady state creep rate as long as the niobium ion is maintained in the Nb⁵⁺ valence state. Material containing 0.4 mol% Nb₂O₅ creeps three orders of magnitude faster than the pure material.Analysis of the results in terms of grain size compensated viscosity suggest that, like “pure” UO₂, the creep rate of Nb₂O₅ doped fuel is diffusion-controlled and proportional to the reciprocal square of the grain size. A model is developed which suggests that the increase in creep rate results from suppression of the U⁵⁺ ion concentration by the addition of Mb⁵⁺ ions, which modifies the crystal defect structure and hence the uranium ion diffusion coefficient.     

A continuum model for cracked UO2 fuel

This paper presents a constitutive model for uranium dioxide fuel pellets in light water reactor fuel rods. The proposed model accounts for the fuel mechanical behaviour under pellet cracking, fragment relocation and pellet-clad mechanical interaction. Moreover, the detrimental effect of cracking on the fuel thermal conductivity is considered in the model. An essential part of the model is the representation of pellet cracks, which significantly affect both the mechanical and thermal behaviour of nuclear fuel under operation. Cracking is modelled in a continuum context, where cracks are represented by nonelastic strains in the material. The continuum representation is particularly suitable for finite element computer codes, since cracking can be treated in the same manner as plasticity and creep. The model is derived in the form of a nonlinear constitutive relation for the fuel material, that may be implemented in either two- or three-dimensional finite element fuel performance computer codes. The fundamentals of the model are presented, and issues concerning its numerical implementation are discussed. The model's ability to capture important aspects of the cracked fuel behaviour is also illustrated by comparisons with in-reactor experiments.     

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.     

Kinetics of initial stage of sintering of UO2 and UO2 with Nd2O3 addition

The kinetics of initial stage sintering of UO₂ powder were reinvestigated, using Ar-10% H₂ atmosphere. The effect of the addition of neodynium oxide was studied. The results revealed that surface and grain boundary diffusion mechanisms act simultaneously. The values of activation energies were found to be $\text{48.48 ± 3.51 kcal/mole}$ in the temperature range 870–942°C and $\text{89.88 ± 9.87 kcal/mole}$ in the temperature range of 942–1030°C for UO₂, and $\text{115.61 ± 7.77 kcal/mole}$ in the temperature range 1030–1150°C for UO₂ + Nd₂O₃. An important decrease in the calculated diffusion coefficient occurs by the addition of Nd₂O₃.     

Corrosion of high burn-up structured UO2 fuel in presence of dissolved H2

The influence of high burn-up structured material on UO₂ corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H₂ for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 × 10⁻¹⁰ M and for Pu at 7 × 10⁻¹¹ M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion.     

Irradiation swelling in UO2-SiO2 fuel under low axial stress

Samples of UO₂-SiO₂ nuclear fuel were irradiated in special capsules, developed for the recording of length changes by very low axial stress (1 kg f/cm²) without any radial restriction. The results of irradiations at temperatures between 200 and 1300°C and burnups up to 20 000 MWd/t are presented. The experimental results emphasize a long-term compaction which is an irradiation-induced creep, and an important swelling occurring at higher burnups for lower temperatures. Between 1100 and 1200°C and at 1 atm this swelling appears at about 10 000 MWd/t.     

Performance of UO2 ceramic fuel in low-power research reactors

The Low Enriched Uranium UO₂ fuel performance in low-power research reactors is assessed in this paper. The usability of this fuel has been demonstrated in some research reactors in the world (SLOWPOKE-2). The fuel proved to be usable in the miniature neutron source low-power research reactors when about 50 fuel rods were substituted by as many dummy rods, while in SLOWPOKE reactors the number of fuel pins reduced by 98. About 3.8531 mk reactivity was rendered available at reactor start-up in MNSRs. The power of MNSRs needed to be increased by about 19%. Shut-down margin, effective shut-down margin, and control rod worth all decreased.     

Reduction of UO2 by H2

The reactivity of H₂ towards UO₂²⁺ has been studied experimentally using a PEEK coated autoclave where the UO₂²⁺ concentration in aqueous solution containing 2 mM carbonate was measured as a function of time at p_H2∼40 bar. The experiments were performed in the temperature interval 74-100 °C. In addition, the suggested catalytic activity of UO₂ on the reduction of UO₂²⁺ by H₂ was investigated. The results clearly show that H₂ is capable of reducing UO₂²⁺ to UO₂ without the presence of a catalyst. The reaction is of first order with respect to UO₂²⁺. The activation energy for the process is 130 ± 24 kJ mol⁻¹ and the rate constant is k_298K=3.6×10⁻⁹ l mol⁻¹ s⁻¹. The activation enthalpy and entropy for the process was determined to 126 kJ mol⁻¹ and 16.5 J mol⁻¹ K⁻¹, respectively. Traces of oxygen were shown to inhibit the reduction process. Hence, the suggested catalytic activity of freshly precipitated UO₂ on the reduction of UO₂²⁺ by H₂ could not be confirmed.     

Fission gas release from UO2 fuel during low-temperature irradiation

A new mathematical interpretation is presented of fission gas release from UO₂ fuel during low-temperature irradiation in terms of a defect trap model and the knock-out process. In the present model it is assumed that gas in an intermediate state exists side by side with the dissolved fission gas and that trapped in bubbles. The present model gives a satisfactory interpretation of the relative proportion of isotopes in the steady state fission-gas release. The dependence of the fission-gas release rate on the fission rate is also interpreted; regimes either proportional to the square of fission rate or proportional to fission rate are predicted, depending on the fission rate interval considered.     

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.     

The movement of lenticular pores in UO2 nuclear fuel elements

A simple theory is developed which describes the motion of lenticular pores in a temperature gradient and takes into account evaporation and condensation rates in the pore surfaces. This mechanism, rather than diffusion within the pore, is likely to govern the pore motion if the pores are filled with helium, and it also gives the experimentally favoured $\text{T}{}^{\mathrm{<mtext>?</mtext><mtext>5</mtext><mtext>2</mtext>}}$ temperature-dependent factor in the pore velocity. Experimental velocities for UO₂ are reproduced quite well by the theory provided that the O/U ratios were slightly greater than 2 in the hot pore region. At temperatures around 2000 K the latter ratio is critical in giving the equilibrium vapour pressure in the pore and hence the pore velocity.     

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.