Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.     

The mechanisms of in-pile fission gas release from UO2

In-pile release mechanisms of fission gas from UO₂ at low temperatures were studied. The release of ¹³³Xe, ¹³⁵Xe, ¹³⁸Xe, ^85mKr, ⁸⁸Kr and ⁸⁷Kr from a sintered UO₂ pellet was measured at temperatures ranging from 250 to 930°C using a graphite specimen holder. The release from the holder, in which a fraction of fission gas was recoil-implanted, was subtracted to obtain the net release from the UO₂ pellet. Knock-out release from the UO₂ was measured directly, and it was found that it was not the main release mechanism, at least not for short-lived nuclides. A ‘pseudo-recoil’ release model is proposed to explain the low temperature release under irradiation. In the model, some of the defects produced by fission fragments act as short-lived carriers for fission gas.     

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

A model for the release of stable fission gases by diffusion from sintered LWR UO₂ fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model.     

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.     

Short-lived fission product release from UO2: Annular versus solid pellets

We have irradiated, in a “dual sweep” test assembly, one element containing solid UO₂ pellets with four, $\text{1 × 1}$ mm surface slots and one element containing annular UO₂ pellets with a central hole two mm in diameter. Data are reported for a linear power range of 17–57 kW/m; thermocouples monitored operating temperatures. He-2%H₂ carrier gas swept short-lived fission products from both elements past independent spectrometers for identification and measurement. The releases from the solid and annular fuel were power-dependent from 17–57 kW/m, with the annular fuel release up to eight times greater than that for solid fuel. The annular fuel also showed a larger released iodine inventory in fuel and sheath areas with access to the carrier gas.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

Behaviour of fission gas products in single crystal UO2 during intermediate temperature irradiation

A new mathematical interpretation is presented of fission gas release from monocrystalline uranium dioxide fuel during intermediate temperature irradiation in terms of a defect trap model, knock-out process and diffusion of bubbles. 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. It is assumed also that the isolated gas atoms, being re-dissolved, are immobile.The present model gives a satisfactory interpretation of the relative proportions of isotopes in the steady state fission gas release at diffrent temperatures. The dependence of fractional fission gas release on fission rate is also interpreted; regimes either proportional to fission rate or inversely proportional to fission rate are predicted depending on the fission rate interval considered. Both temperature dependent and temperature independent fission gas release can arise.The presented dynamic method of studying the release of fission gases during irradiadion provides a further test beside the static method of the veracity of the assumed mechanisms. Calculations show that fission gas behaviour becomes more complex for oscillated fission rate in the regime where the fractional release is inversely proportional to the fission rate for the steady state.     

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.     

Re-solution effects and fission gas swelling in UO2

A fission gas swelling model is proposed which enables one to calculate swelling in the vicinity of grain boundary networks and in imperfection-free regions. The grain boundary swelling requires a knowledge of the gas accumulation and the reaction-rate at the boundary. The gas accumulation was calculated by deriving a modified form of Fick's second law wherein it was assumed that because of re-solution effects the in-pile diffusion coefficient can be described as a function of the gas concentration but is independent of the actual operating time. Reaction-rates for bubbles at grain boundaries were derived in the manner discussed by de Jong and Koehler in their treatment of vacancy clustering. The results indicate that there is a grain size of about 10⁻⁴ cm for which the swelling is a maximum, which increases somewhat with irradiation temperature and with depletion at a constant temperature. The results enable one to predict the swelling and the mean radii of both intergranular and intragranular bubbles. Mean bubble radii predicted using the re-solution swelling model are in reasonable agreement with radii obtained from electron micrographs of irradiated UO₂ fuel samples. It is argued that gas bubble migration is the predominant means by which gas atoms arrive at grain boundaries at irradiation temperatures above about 900°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.     

UO2 dissolution in Boom Clay conditions

The solubility of uranium dioxide (UO₂) was measured in real and synthetic Boom Clay waters with varying concentrations of humic acids and carbonate under reducing conditions at 20 °C. Uranium concentrations in function of time suggest the reduction of U(VI) to U(IV) by the humic acids which is occurring faster in real clay water than in synthetic clay waters. Humic acids induce also a competition to complex U(VI) in carbonate-containing solution, but they are not able to control the uranium concentration at high bicarbonate concentration (0.02 mol dm⁻³). Nevertheless they may play a role at low carbonate concentration. In our experimental conditions, the geochemical calculations indicate that two uranium secondary phases (U₄O₉ and UO_2(c)) are susceptible to control the uranium concentration in solution. These calculations are in good agreement with results of the X-ray photoelectron spectroscopy. At the end of tests, uranium concentrations reach steady-state values between 3 × 10⁻⁸ and 5 × 10⁻⁸ mol dm⁻³ in the bicarbonate-rich solutions. Although these concentrations are considered as conservative, they are 10-100 times higher than in natural Boom Clay. The consequence is that spent fuel could slowly dissolve in the interstitial clay water undersaturated with respect to UO₂/UO_2+x of the fuel.     

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.     

The development of a mechanistic code on fission product behavior in the polycrystalline UO2 fuel

The current status of a mechanistic code (RTOP) on fission product behavior in the polycrystalline UO₂ fuel is described. Outline of the code and implemented physical models is presented. The general approach to the code validation is discussed. It is exemplified by the results of validation of the models of oxidation and grain growth. The different models of intragranular and intergranular gas bubbles behavior have been tested and the sensitivity of the code in the framework of these models has been analyzed. An analysis of available models of the resolution of grain face bubbles is also presented.     

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.     

UO2燃料裂变气体渗流模型研究

为分析UO₂燃料晶界气泡连通导致裂变气体间歇性释放的动力学过程,从而解决目前扩散模型预测的沿芯块径向释放份额与实验测量不符的问题,采用二维渗流模型模拟UO₂燃料晶界气泡网络的演化及与燃料棒内自由空间连通的释放过程。研究结果表明,渗流模型预测沿芯块径向的裂变气体释放份额在芯块中间部分出现局部峰值,并随着时间向芯块外侧推进,与辐照试验观察到不同燃耗下径向裂变气体分布现象定性符合。因此,本研究建立的渗流模型能够从机理上解释此前扩散模型未能预测的UO₂燃料裂变气体释放份额沿径向非单调分布现象。