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.     

The use of U3Si2 dispersed fuel in Low-Power Research Reactors

The use of U₃Si₂ as a Low Enriched Uranium (LEU) dispersed fuel in Low-Power Research Reactors is investigated in this paper. The fuel proves to be usable if some of the original fuel rods (HEU UAl₄–Al fuel) are still simultaneously employed (mixed core) without changing the structure of the actual core. About 3.5712 mk Initial Excess Reactivity (IER) is procured. Although the worths of both the control rod and the reactivity devices decrease, the safety of these reactors is higher in the case of the new LEU fuel. If the dimensions of the meat and/or the clad are allowed to change these reactors can be run with a meat 2.15 mm outer radius, and a clad 0.58 mm thickness. The IER will then be 4.1537 mk, and both the control rod (CR) worth and the safety margins decrease.     

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.     

An experimental study of the shielding characteristics of the dwelling house building materials against gamma radiations in the Central Region of Syria

The shielding properties of the concrete and blocks used for the construction of dwelling houses in the Central Region of Syria (CRS) were measured and studied. The concrete used for the ceiling construction was found to have optimum shielding properties with 0.182 cm⁻¹ (or equivalently 0.0859 cm² g⁻¹) for the linear (mass) attenuation coefficient [L(M)AC]. In addition gamma radiation is attenuated by 73.221% on average, while the blocks used for the walls have smaller LACs (0.082 cm⁻¹ for the bare blocks, and 0.118 cm⁻¹ for the coated ones). Although the LACs for the blocks are smaller than those for the concrete their shielding properties are good to protect from the gamma radiations coming from radioactive or nuclear accidents (78.630% attenuation), even Chernobyl – like disasters, because of their big width (10–12 cm). The LACs were measured by an ionization chamber and simple theoretical calculations have been made to predict the concrete LACs. The calculations showed an average LAC for the six samples equal to 0.1664 cm⁻¹ with 8.47% error with respect to the experimental values.     

New models for UO2 fuel structure evolution under irradiation in fast reactors

On the base of analysis of experimental observations and critical assessment of existing models for oxide fuel structure evolution under operation conditions of fast reactors, new models for fuel restructuring and coring are proposed. The restructuring model describes coherent motion in the temperature gradient of various voids (gas bubbles, sintering pores and large lenticular pores) and grain boundaries, to which the voids are attached. As a result, the model explains elongation of thermally growing equiaxed grains and formation of columnar grains, and predicts a rapid formation of extended columnar grain zone during a relatively short initial period of fast reactor irradiation. The coring model describes formation and growth of the central void in the fuel pellet, activated by mass transport from the inner to the outer zone of the pellet under stresses induced by inhomogeneous fuel densification in the initial period of irradiation.     

Some observations of density and porosity changes in UO2 fuel irradiated in water-cooled reactors

A technique has been developed for the hot-cell measurement of the apparent density of irradiated UO₂ fuel after extraction from a fuel pin. A single determination is accurate to $\text{± 3 \%}$ at the 95 % confidence limit. The method has been applied to fuel irradiated in thermal neutron fluxes in the Winfrith SGHWR and in the Halden BWR. Material has been examined at ratings of 1–51 W/g and in the burn-up range $\text{0.09–5.79 × 10}{}^{20}\text{fissions/cm}$. It is concluded that pellets with peak temperatures below 1100°C densify during irradiation, but at higher temperatures the pellets begin to swell. Fuel micrography has shown that the densification is principally due to the loss of micropores with a temperature dependency given by an activation energy of 5200 cal/mol. Above 1000°C the densification is masked by the formation and growth of intergranular fission gas bubbles, whose volume may exceed that of the manufactured pores which have sintered. In solid fuel pellets central swelling did not balance densification in the cooler rim until the fuel centre temperature exceeded 1700°C.     

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.     

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.     

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.     

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.     

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.     

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.     

Remarks on the sintering behavior of UO2-Gd2O3 fuel

The previous work by our group showed experimental evidence that supports the idea that a diffusion barrier is formed around Gd₂O₃ agglomerates due to the formation of gadolinium-rich (U,Gd)O₂ phases with low diffusivity. This would be the reason for the bad sintering behavior of the UO₂-Gd₂O₃ fuel. The objective of this investigation was to confirm that hypothesis by direct experimental evidence. Analysis of the results showed that the diffusion barrier hypothesis is not applicable.     

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.     

Temperature profiles in UO2 fuel under direct electric heating conditions

The thermal environment of UO₂ fuel in reactor may be simulated under conditions of direct electric heating (DEH). Using reasonable assumptions, the complex thermal/electrical system is modeled mathematically by the DEHSSTD code. The algorithm computes the thermal and linear power profiles in the fuel and in addition a history dependent scale factor for the thermal conductivity and for the electrical conductivity. These account for the dependence of materials properties on the physical state of the fuel. The DEHSSTD model is applicable under both steady-state and transient conditions. Convergence of the DEHSSTD model is studied and optimization performed for the model's fuel parameters. DEH and reactor thermal environments are compared, the DEH temperature profile being more sharply peaked at the fuel's center. An equivalent nuclear power is defined on the basis of the DEH temperature distribution.     

Phasengleichgewichte in den systemen UO2-UO2,67-ThO2 und UO2 + x-NPO2

Solid-state chemical investigations have established that in the compositional range UO₂-UO_2.67-ThO₃ of the U-Th-O ternary system, the following single-phase domains exist: U₃O₈, which does not dissolve any ThO₂ in the solid state; an ordered M₄O₉ phase on the section between U₄O₉ and U₂Th₂O₉, below ≈ 1150 °C; and a phase with fluorite structure which occupies a large part of the system and which at 1250 °C is bounded by the compositions UO₂-UO_2.25 (U_0.43, ThO_0.57)O_0.25-ThO₃. The maximum O/M ratio of the “fluorite” phase is O:(U + Th) = 2.25. The highest oxidation valency of uranium is 5.30; this value falls as more thorium oxide is incorporated in the (U.Th)O_{2 + x} “fluorite” phase.