The salient features of using a solid substance to cool the core of a nuclear reactor and the associated advantages and limitations are examined. Conceptual proposals concerning the core design and the arrangement of the in-reactor space of a high-temperature nuclear reactor with a solid coolant are presented. Evaluated data and some results for a model reactor are presented.The development of nuclear power requires an examination and development of innovative reactor designs which have enhanced safety due to the inherent properties of a reactor and are highly cost-effective. One interesting and promising proposal is to use a solid substance as the coolant [1].There are substantial advantages to using a solid coolant to cool the core of a nuclear reactor. These include the possibility of the coolant moving in the core under gravity and the absence of any need for excess pressure in the vessel. In turn, this means that the metal content of the system is low, the risk of accidents is lower, and the scale of the consequences of accidents is smaller. The core of a reactor can be cooled with a solid substance when certain conditions associated with the characteristic features of the solid coolant are satisfied. The most important requirements are uniform continuous motion of the coolant with minimum density fluctuations in all sections of the core, high mechanical strength and durability of the particles, and good heat-transfer indicators, i.e., the coolant material must have a high thermal conductivity and heat capacity under the working conditions characteristic for the core of a nuclear reactor.Studies of the possibility of using a solid coolant based on finely dispersed particles of graphite for cooling the core of a thermal reactor have revealed concrete conditions which are necessary in order to satisfy the main requirements for a solid coolant. A brief report on the results of such investigations is given in [2]. To prove that the proposed coolant can move under gravity as a dense layer with a high velocity and to study the durability of the particles, a complex of experimental works was conducted at the Research Institute and Scientific-Industrial Association Luch. The results of these investigations confirmed that a solid coolant consisting of spherical graphite particles with average diameter 1 mm coated with a pyrolytic carbon coating can move uniformly under gravity. When such a coolant is used, coolant velocities and heat-transfer coefficients which make it possible to obtain energy release density characteristic for high-temperature gas-cooled reactors can be

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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.