K. Lassmann

TRANSURANUS: a fuel rod analysis code ready for use*

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

The radial distribution of plutonium in high burnup UO2 fuels

Abstract A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21000 and 64000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions.

Modelling the high burnup UO2 structure in LWR fuel

The concept of a burnup threshold for the formation of the high burnup UO2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60–75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U.

Oxide fuel transients

Abstract The behavior of UO 2 fuel from power reactors has been studied up to high burn-up (~ 60 GWd/tU), under both steady state and power transient conditions in the range of 35 to 48 kW/m. Careful post-irradiation examination involving electron probe microanalysis, transmission and scanning electron microscopy in addition to detailed hot cell examination have provided a large data base on the radial migration and release of volatile fission products. Theoretical evaluation with different computer codes and supporting laboratory experiments established the basis for understanding the kinetics and mechanisms operative during transients in high burn-up UO 2 . Typical examples are given to demonstrate the degree of understanding achieved

Numerical algorithms for intragranular fission gas release

Complicated physical processes govern fission gas release in nuclear fuels. Besides the physical problem, there is a numerical problem since some solutions of the underlying diffusion equation have numerical errors that by far exceed the physical details. In this paper, the efficiency and the accuracy of some numerical solutions are analysed. Random operation histories were generated and the errors inherent in each algorithm evaluated over a wide range of up- and down-ramps by comparing the results with the quasi-exact ANS-5.4 algorithm. The URGAS algorithm can be considered as well balanced over the entire range of fission gas release. The new FORMAS algorithm is superior at fission gas release above f≈0.05 and may in a physical sense be considered as an exact solution in this range. Unfortunately, the deficiency of this most elegant and mathematically sound algorithm at low fission gas release could not be fully overcome. However, in view of the many inherent uncertainties, both algorithms are considered as sufficient to be used in a fuel rod performance code. All algorithms analysed in detail can be made available on request as FORTRAN subroutines.

The oxired model for redistribution of oxygen in nonstoichiometric uranium-plutonium oxides

Abstract The oxygen distribution model OXIRED is described which predicts the steady-state and transient radial oxygen-to-metal ratios in a fuel rod as a function of the average burnup, the radial temperature profile and the time for hyperstoichiometric oxides UO 2 + x (LWR conditions) and hypostoichiometric mixed oxides (U 1−q Pu q )O 2−x (FBR conditions). Numerous calculations with the fuel rod performance code TRANSURANUS into which the OXIRED model has been incorporated have proven that the newly developed numerical algorithm is extremely fast and reliable.

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

Abstract The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235 U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

Oxygen potential measurements in irradiated mixed oxide fuel

In an irradiated (U, Pu)O2 ± x fuel, radial oxygen redistribution occurs within the temperature gradient. Electromotive force measurements on drillings from different radial positions of a quenched pin show that the shape of the O/M-profile is not in agreement with that expected from models and theory. Previous determinations by other authors using electron-microprobe or X-ray measurements, however, yield profiles with shapes similar to that observed here. These profiles are characterized by a steep decrease in O/M in the center of the fuel. A transient O/M-redistribution model was established yielding time constants for redistribution during cooling. The numerical cases calculated show that the quenching procedure does not affect the O/M-profile. Though each of the above three techniques used to experimentally determine O/M-profiles on irradiated oxide pins has its shortcomings, the totality of the results is suggested to contain enough evidence to rediscuss and reconsider the basic assumptions and mechanisms of oxygen redistribution in (U, Pu)O2 fuels.

Fission gas and caesium gradients in single grains of transient tested UO2 fuel: Results of an EPMA investigation

Abstract Measured concentration profiles for retained xenon and caesium in single grains located in several different radial positions are presented. Effective diffusion coefficients derived from the retention data varied from 10−15 to 10−10 cm2 s−1 depending on the radial location of the grain. Caesium release began at a higher temperature than xenon release and at a given location D ∗ Cs was generally about a factor of 2 lower than D ∗ Xe .

A New Data-Condensation Method Based on Multidimensional Minimisation.

Abstract During irradiation in the Halden Reactor data such as linear rating, coolant temperature, or reactor state are routinely stored every 15 minutes, in the case of special events even at shorter time intervals. Because of the long irradiation times of up to several years, a huge amount of data is produced which is considered to be too much for direct input into fuel performance codes. In order to reduce the amount of data, various data-condensation procedures have been developed, which are briefly discussed. An innovative data-condensation method based on multidimensional minimisation is presented. The condensation factor obtained so far for Halden irradiations is 25–40, which is sufficient for a modern fuel rod performance code like TRANSURANUS. This new condensation method avoids all loss of experimental data and doubtful averaging of temperatures associated with other more drastic condensation methods. Up to now, the so-called reactor state variable, which defines the reactor operation (e. g. constant, decreasing or increasing power), has not been used. The introduction of this variable should further improve this new condensation method.