A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

Abstract

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to lowenriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

Innovative and safe supply of fuels for reactors

“…The very first in-pile tests (IRIS2, FUTURE, IRIS3 0.3%Si) of nuclear fuels with a UMo/Al composition showed an unacceptable swelling under irradiation, in some cases even leading to plate breakaway, even though these tests were only performed with limited surface power ( 350 W.cm À2 ) [6][7][8]. The failure has been traced back to a UMo/Al Inter-Diffusion Layer (IDL) growing during in-pile irradiation at UMo-Al interfaces and to its unsatisfactory properties under irradiation [9].…”