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A validation of the possibility of developing and the basic advantages of a high-temperature nuclear reactor where the first-loop coolant is a solid are presented. The basic requirements for a solid coolant are formulated, a technology for fabricating spherical graphite particles by gas-phase pyrolytic deposition is developed, and three experimental batches are prepared. The experimental facilities for investigating the motion and heat transfer, including coolant flow stability, heat exchange, and durability, are described. The results of a determination of the heat-emission coefficient during the flow of the solid coolant in a 10-mm in diameter circular channel with warming-wall temperatures in the range 373-1073 K and flow velocities 0.1-0.22 m/sec in vacuum, argon, and helium are presented. The requirements for a 500-kW bench model, on which the basic parameters of the nuclear power system with a solid coolant are to be obtained, are formulated.If nuclear reactors are viewed as the main source of electricity and heat for purposes of global conservation of gas and oil and decreasing emissions of combustion products into the atmosphere, at least three problems must be solved:• the possibility of accidental emission must be eliminated and the escape of nuclides into the biosphere during operation of the entire chain of the nuclear cycle must be decreased; • the technological schemes for using the stores of nuclear fuel, which allow for economically competitive large-scale operation of nuclear power for a long period of time, must be validated; and • the possibility of using the nuclear components of power systems for producing nuclear weapons must be eliminated. The improvement of the currently widely used PWR is making these reactors more complicated and expensive, but nevertheless the probability of meltdown of the core of such a reactor is estimated to be 10 -5 yr -1 and the probability of radionuclides escaping over the last barrier in the form of a closed shell is 10 -7 yr. Estimates show that more than 10000 nuclear reactors will be needed to meet the world-wide demand for electricity. In this case, the probability of a serious accident with radionuclides escaping would be 10 -7 ·10 4 = 10 -3 , which is unacceptable for mankind. Accidents with escape of radionuclides beyond the last barrier must be eliminated if nuclear energy is to be used on a large scale.

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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

The concept of a direct-flow channel reactor with supercritical-pressure water (CR-SCP) is presented. Neutron-physics, thermohydraulic, and strength calculations are used to substantiate the fundamental core design with a heavy-metal moderator which at supercritical pressure is competitive with other modern reactor designs with respect to fuel-cycle indicators. Two types of fuel-element and fuel-channel structures are examined. It is shown that fuel elements based on micropellets and a metal matrix are highly reliable and have higher operating characteristics (burnup, service life, geometric stability, and so on) than fuel elements with uranium-dioxide fuel. A CR-SCP design and the technological scheme of a power-generating unit are presented, and the systems which are required to ensure normal operation and safety are determined. The main technical-economic indicators of a power-generating unit with installed electric power 850 MW are estimated.The interest in supercritical coolant pressure in reactors is due to the fact the new reactors must be competitive. In the last few years a surge of interest has been observed in Japan, USA, Germany, France, Canada, and other countries [1]. Reactors with supercritical-pressure water are being considered as one direction in the international program for the development of fourth-generation reactors.Some countries are undertaking efforts to develop direct-flow water-moderated water-cooled vessel reactors with supercritical-coolant pressure. The core of these reactors can be designed to operate on thermal and fast neutrons with almost the same thermal layouts of the power-generating units and the same efficiency. It is important to note that a fast-neutron reactor makes it possible, aside from appreciably decreasing the capital costs, to improve fuel utilization [2], since it makes it possible to increase the breeding ratio of nuclear fuel up to 1 and aim at a closed fuel cycle. However, such reactors are characterized by a change in water density by a factor of 10 from the core entrance to the core exit. This is accompanied by a change in the neutron spectrum over the height of the core and makes it difficult to smooth and stabilize the energy-release fields not only