The increasing safety requirements for nuclear power systems are determining the direction for developing systems that would increase the self-protection of reactors. This problem is urgent for water-graphite channel reactors with a positive loss-of-water effect. The use of absorbers -boron, erbium, hafnium, and gadolinium -decreases loss-of-water in RBMK [1]. Gadolinium possessses a high microscopic neutron absorption cross section in the thermal range, so that as a rule low concentrations are used as a consumable absorber to compensate for fuel burnup during a run. In this case, the loss-of-water effect in uranium-graphite reactors changes very little. In contrast to gadolinium, erbium possesses a strong resonance in the thermal range (~0.5 eV). In this case, when water is lost from the core the thermal-neutron spectrum becomes becomes a high-energy spectrum and, as a result, the resonance absorption of thermal neutrons by erbium increases appreciably. Thorium possesses a small cross section for radiative capture in the thermal range, so that when thorium is used as an absorber the loss-of-water effect decreases as a result of an increase in the concentration and heterogeneous arrangement in a fuel assembly.The advantage of using thorium as an absrober lies in the fact that the loss-of-water effect decreases and breeding of nuclear fuel occurs in a manner so as to maintain the required excess reactivity during a run. A fuel assembly includes sleeve fuel elements with bilateral cooling and thorium-absorbing absorbing elements which are placed in the central cavity. The latter elements can be made in the form of either short rods, which are strongly secured in each fuel element, or a long continuous rod [2,3]. A fuel-assembly with the short-rod variant of an absorbing element makes is easier to load the element into and unload it from a reactor. The presence of a long rod increases the number of degrees of freedom for ensuring reactor criticality and safety. To prevent the absorbing elements from touching the inner surface of a fuel element a fuel assembly can be equipped with spacing inserts or ribs. This type of construction of a fuel assembly requires neutron-physical and thermohydraulic optimization.The purpose of optimization is to determine the geometric dimensions and characteristics of a fuel assembly that would maintain the required reactivity excess and degree of subcriticality during a run and make it possible to obtain a negative loss-of-water effect. Our objective in the present article is to estimate the thermohydraulic parameters that would make it possible to achieve reliable and efficient heat removal from the elements of a fuel assembly under normal operating conditions and a small coefficient of hydraulic resistance of the inner ring-shaped gap between an absorbing element and a sleeve fuel element as compared with the outer ring-shaped gap between a fuel element and a channel, so that in the case of a loss-of-coolant accident water would be rapidly removed from the inner ring-shaped gap. Water...

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.