P. Van Uffelen

On the role of grain boundary diffusion in fission gas release

Abstract It is generally believed that thermal fission gas release from LWR fuel occurs mainly via interconnected grain boundary bubbles. Grain boundary diffusion is not considered to be a significant mechanism. We investigated this supposition by two methods; first, by assessing the distance a gas atom can migrate in a grain boundary containing perfectly absorbing traps. For areal number densities and fractional coverages by the traps observed in fuel irradiated to burnups exceeding ∼20 MWd/kg, gas atoms will be trapped after a migration distance equal to the size of a grain or less. This supports the supposition for medium-to-high burnups. However, the above-mentioned model is inapplicable for trace-irradiated specimens. In our second analysis, we examined Xe release from trace-irradiated UO 2 . The measurements indicated that the liberation involves more than only lattice diffusion at the specimen surface, and that the data are consistent with sequential lattice and grain boundary diffusion unimpeded by intergranular traps. The analysis also provided rough estimates of the grain boundary diffusion coefficient in UO 2 .

Modelling thermal conductivity and self-irradiation effects in mixed oxide fuels

The present paper describes a physical model for thermal conductivity of actinide oxides. The model is based on the Debye-Einstein theory of thermal energy of ionic dielectrics, on the Klemens-Callaways approach for the heat conductance modelling and on correlations between thermoelastic properties of solids. Some results of calculations and a comparison between the calculated and measured values of the thermal conductivity of UO 2 , UO 2 -Gd 2 O 3 and also of ThO 2 , AmO 2 , Am 2 O 3 and Cm 2 O 3 , for which a limited data-set is available in the open literature, are reported. Moreover, self-irradiation effects in AmO 2 , Am 2 O 3 , (Am,U)O 2-x , (Am,Np,U)O 2-x and Cm 2 O 3 were analysed with the developed model.

MULTISCALE MODELLING FOR THE FISSION GAS BEHAVIOUR IN THE TRANSURANUS CODE

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of UO₂ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

High frequency acoustic microscopy for the determination of porosity and Young's modulus in high burnup uranium dioxide nuclear fuel

During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of the porosity and of elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. Ultrasound waves are generated by a piezoeletric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A 67 MWd/kgU UO2 pellet was characterized using the acoustic microscope installed in the hot cells of the Institute of Transuranium Elements: 90 MHz frequency was applied, methanol was used as coupling liquid and VR was measured at different radial positions. By comparing the porosity values obtained via acoustic microscopy with those determined using ceramographic image analysis a good agreement was found, especially in the areas close to the centre. In addition Youngs modulus was calculated and its radial profile was correlated to the corresponding burnup profile.

Two-Way Coupling Between the Reactor Dynamics Code DYN3D and the Fuel Performance Code TRANSURANUS at Assembly Level

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models.A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states.Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up.The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.Copyright

3.19 – Oxide Fuel Performance Modeling and Simulations

The ultimate goal of simulating nuclear fuel is to predict a fuel rod’s behavior and lifetime in a reactor. Doing so requires taking into account the coupled effects of heat transfer, the mechanical interaction between the fuel and its surrounding protection, the isotopic evolution caused by the irradiation, and the chemical interactions between fuel, fission products, cladding, and coolant. In view of the strong interactions between the various aspects of fuel performance and the resulting mathematical problems, the simulation of the fuel behavior requires fuel performance codes, which are being used by safety authorities, research organizations, and fuel vendors. The basic equations for each aspect of the codes are described in this chapter, together with the specific features of and modeling requirements for the design basis accident analysis. Finally, this chapter also outlines the advanced issues and future needs of fuel performance codes for nuclear fuel rods based on oxides.

Modelling the variable precipitation of fission products at grain boundaries

We have developed a model for the precipitation of fission products in a grain boundary which embodies a variable reaction rate on the precipitate surface. This enables us to account for modifications of the local fuel chemistry, or to distinguish between the behaviour of different migrating species. In addition, we have assessed the influence of the trapping parameters on the precipitation rate according to different models from the open literature which have been extended in order to incorporate the variable intrinsic reaction rate. The interrelationships among the models have been established while their limitations and range of validity have been discussed. The results reveal that there is a critical value above which the influence of the intrinsic reaction rate, between a fission product and an intergranular trap, on the global precipitation rate becomes negligible.

High temperature measurements and condensed matter analysis of the thermo-physical properties of ThO2

Values are presented for thermal conductivity, specific heat, spectral and total hemispherical emissivity of ThO2 (a potential nuclear fuel material) in a temperature range representative of a nuclear accident - 2000 K to 3050 K. For the first time direct measurements of thermal conductivity have been carried out on ThO2 at such high temperatures, clearly showing the property does not decrease above 2000 K. This could be understood in terms of an electronic contribution (arising from defect induced donor/acceptor states) compensating the degradation of lattice thermal conductivity. The increase in total hemispherical emissivity and visible/near-infrared spectral emissivity is consistent with the formation of donor/acceptor states in the band gap of ThO2. The electronic population of these defect states increases with temperature and hence more incoming photons (in the visible and near-infrared wavelength range) can be absorbed. A solid state physics model is used to interpret the experimental results. Specific heat and thermal expansion coefficient increase at high temperatures due to the formation of defects, in particular oxygen Frenkel pairs. Prior to melting a gradual increase to a maximum value is predicted in both properties. These maxima mark the onset of saturation of oxygen interstitial sites.