Neutron-physical and heat-engineering aspects of uranium-graphite reactors loaded with enriched-uranium fuel elements

Abstract

The changes proposed in the fabrication technology for enriched-uranium fuel elements to be used in uranium-graphite reactors are associated with changes made in the fuel-element construction which could influence the local uranium burnup characteristics and the axial and radial distribution of the neutron flux density over the fuel elements and the core, the energy distribution in a fuel element and the process channel, heat transfer, and possibly the decrease of margin up to boiling. These aspects are all important for the heat-engineering reliability and operational safety of reactors.Enriched-uranium fuel elements are disperion-type fuel elements [1]. They consist of a cylindrical uranium-dioxide kernel with an aluminum base in a sealed aluminum-alloy cladding. The fundamental difference between fuel elements with this construction and the old-type fuel elements is that the height of the kernel is smaller while the height of the fuel element itself is unchanged (Fig. 1).In a tansverse section, the process channel consists of a cylindrical tube with five longitudinal ribs along the inner surface to prevent the fuel elements from touching the wall. Coolant for removing heat is fed into the gap between the outer surface of a fuel element and the inner surface of the channel. The moderator is graphite. The graphite masonry consists of graphite blocks with process channels, loaded with fuel elements, arranged vertically in the central opening.The neutron spectrum and the axial distribution of the neutron flux over the kernel and the cladding of the fuel elements must be determined in order to estimate the changes in the thermophysical parameters of the coolant. A series of calculations was performed, using the SCALE three-dimensional computer program system [2] and a model cell, including a graphite block, a process channel, the coolant (water), and a fuel element (Fig. 2), to solve the neutron-physical problem. A fuel element can be repreented in the axial direction by three components: the kernel and the top and bottom ends of the cladding, which are divided in the vertical direction into ten and three, respectively, equal zones. The 235,238 U nuclei are undiformly distributed along the entire volume of a kernel. The mass of the fissioning isotopes and the radius of the kernel are identical for both types of fuel elements.To deteremine the axial distribution over the fuel elements, we shall use a cell loaded only with enriched-uranium fuel elements. The eccentricity of all fuel elements is assumed to be zero.The following approximations were introduced for the model:• the absence of ribs is compensated by an effective addition δ pc to the process-channel wall (Fig. 3); estimates show that switching to such a model does not result in any appreciable error in the calculations of an elementary cell of this type for enriched-uranium fuel elements; • in the calculations pure aluminum is used as the structural material for the process channel and the kernel of the fuel elements.