Patent Number: 040452867
Section: summary

This invention relates to a nuclear reactor in which the fuel is provided in the form of a molten salt and is more particularly concerned with an arrangement of the reactor block which serves to limit to strictly necessary values both the length and bulk of the circuit followed by the high-temperature molten salt which is discharged from the reactor core and consequently to reduce the thermal stresses and chemical corrosion of metallic components employed in the construction of the reactor. The conceptual design and technology of molten salt reactors are already known. These reactors make use of a fuel in liquid form which is brought to a high temperature of the order of at least 600.degree. C. as a result of nuclear fission within the reactor core. As a general rule, this fuel consists of plutonium or uranium fluoride or alternatively a mixture of uranium and thorium fluoride dissolved in fluorides of lithium-7 and beryllium; the eutectic mixture thus formed consequently has a relatively low melting point, suitable fluidity and low vapor tension. In a reactor of this type, provision is made within the core for a mass of suitable neutron-moderating material usually consisting of graphite in which are formed ducts for the flow of the fuel salt; the heat gained by this latter as it passes through the reactor core is exchanged in at least one primary heat exchanger with another molten salt or so-called buffer salt such as sodium fluoborate, for example. Said buffer salt in turn exchanges its heat in a secondary circuit comprising a steam generator, the steam being finally expanded in an electric power generating plant. Reactors of this type are capable of operating with a flux of thermal neutrons or a flux of fast neutrons according to the composition of the salt, the distribution of the salt within the reactor core and the nature of the moderator. The fuel salt can be burnt in the reactor core with a periodic readjustment of the fuel concentration whilst processing of this latter and in particular the extraction of fission products are carried out only after a predetermined period of operation. In another design concept, the fission products and especially the gaseous products are continuously withdrawn by a method of chemical bubbling with concomitant adjustment of the fuel salt concentration. Finally, in the case of operation as a breeder reactor, the fuel salt is processed so as to permit continuous removal of protactinium-233 by liquid bismuth followed by metallic reduction by thorium and intermediate storage so as to permit radioactive decay and conversion to uranium-233 in the state of fluoride which is then returned to the main primary circuit (Revue Energie Nucleaire -- vol. 13 No 2 -- March, 1971 -- "Les reacteurs a sels fondus" -- (Molten-salt reactors) -- M. Grenon and J-J. Geist). In these conventional design concepts, the three essential components of the primary molten fuel-salt circuit, namely the reactor core, the primary heat exchangers and the pumps for circulating the fuel within said circuit are connected to each other by means of piping systems forming one or a number of loops located outside the vessel which contains the reactor core. These piping systems must have a sufficient degree of flexibility in order to maintain thermal stresses at an acceptable level. Moreover, in designs of this type which are at present known, the pumps are placed either in the hot branch of said loops for collecting the molten fuel salt at the outlet of the reactor core or in the cold branch at the outlets of the primary heat exchangers. In point of fact, these solutions have a disadvantage in that a substantial portion of the primary circuit is placed in contact with the fuel salt at its maximum temperature; this accordingly produces not only an increase in thermal stresses but also an aggravation of the problems presented by chemical corrosion of structures by the molten salt since the corrosive action of this latter increases very rapidly as the temperature rises. The present invention relates to a nuclear reactor of the type aforementioned in which the primary circuit is directly integrated in the vessel containing the reactor core, thus circumventing the disadvantages outlined in the foregoing. In particular, said primary circuit is provided within a common vessel containing the reactor core and a neutron-moderating mass pierced by passages for the circulation of the molten fuel salt with at least one primary heat exchanger which is located as close as possible to the reactor core and through which the hot fuel salt passes immediately after discharge from said core and with pumps for circulating the cold molten fuel salt which is discharged from the heat exchangers and returned into the reactor core. In accordance with the invention, the molten fuel salt reactor of the type described above is distinguished by the fact that the free spaces defined within the reactor vessel between the core, the heat exchangers and the pumps are filled with an inert material which is compatible with the molten fuel salt except for the passages in which the fuel salt is circulated. The arrangement adopted consists especially in mounting within a common vessel both the reactor core, the heat exchangers and the pumps for circulating the fuel salt which is discharged from said heat exchangers and returned to the core. Accordingly, the useful volume of the molten fuel salt which is heated to the maximum temperature of the cycle can be reduced to the strict minimum; the primary heat exchanger or exchangers or the separate units constituting said exchangers can be connected directly to the core outlet by means of passages of small dimensions while the assembly formed by the reactor, the heat exchangers and the pumps is placed within a single container or vessel which surrounds the entire primary circuit. In consequence, the metallic structures or other structures which are subjected to the most arduous operating conditions and placed in particular in contact with the hot fuel salt are limited to the greatest possible extent. Thus most of the primary circuit is only in contact with the cold fuel salt for which connecting passages are also provided, the dimensions of said passages being calculated as a function of the requirements of hydraulic operation of the system. Moreover, the integrated reactor concept results in containment of the molten fuel salt within a vessel of simple shape which is conducive to cooling and heat insulation. In a particular embodiment of the invention, the reactor core is placed within the central portion of an open vessel having a vertical axis and is surrounded by a lateral reflector which defines an annular region with the internal vessel wall, the pumps for the circulation of molten fuel salt and the heat exchangers being placed within said annular region, said pumps and said heat exchangers being disposed at uniform intervals around the reactor core. In accordance with a particular feature of the first embodiment aforementioned, the pumps and heat exchangers are suspended within the annular space beneath a horizontal vault roof extending above the reactor vessel, said vault roof being provided with a central access opening placed opposite to the reactor core and closed by a removable shield plug. In another alternative embodiment which permits a further reduction in overall length and dimensions of the connecting passages provided for the molten fuel salt discharged from the reactor core, each circulating pump is directly mounted beneath a heat exchanger within the annular space so as to constitute a pump-exchanger unit, the passages providing a connection with the reactor core being constituted by ducts formed radially from the axis of the reactor vessel and placed in the top and bottom portions of the reactor core. Whatever design may be adopted for the primary circuit and whatever in particular may be the relative arrangement of the components of said circuit, the inert material which is compatible with the molten fuel salt and fills the spaces left free between the reactor core, the heat exchangers and the pumps in order to limit the useful volume in circulation is constituted by expanded graphite. The use of graphite within a molten-salt reactor vessel has already been contemplated. However, in known designs, this material is provided in the form of impregnated blocks but this is attended by two disadvantages: on the one hand, said blocks are costly and, on the other hand, they result in local temperature rises which are liable in some cases to be prohibitive. In fact, when the free spaces formed for example within the annular region around the reactor core between the pumps and the heat exchangers are filled with graphite blocks, it is not possible to prevent the presence of thin layers of fuel salt which lie stagnant between the adjacent faces of said blocks, especially as a result of clearance-spaces formed at the time of assembly and operation. Under these conditions, the nuclear components of said salts which are exposed to the environmental neutron radiation give rise to nuclear reactions, thus releasing thermal energy which cannot be removed by conduction through the graphite since this latter usually has low heat conductivity. In some cases, the temperatures attained can be of a high order and prove detrimental to the good operation of the installation since they are liable to result in serious damage to some reactor vessel structures or the internal components of said vessel. The utilization of expanded graphite makes it possible on the contrary to overcome these disadvantages. This graphite is preferably obtained from grains of lamellar complexes of graphite abruptly subjected to a substantial temperature rise so as to produce a thermal shock which results in the conversion of said grains to flakes. Said flakes are then compacted so as to form blocks or graphite agglomerates having a density which can range from 0.1 to 2 according to the compacting pressure adopted. In particular, the compacting can be carried out by isostatic or unidirectional compression, depending on the nature of the end product to be obtained and the design of the fabrication means employed. One remarkable advantage of expanded graphite results from the possibility of forming lightweight compact blocks or masses, the faces of which are practically impermeable to liquids which have a high surface tension. This is precisely the case of the molten fuel salt which is circulated within the reactor in contact with said blocks and fills the free spaces in the reactor vessel. It is worthy of note that these free spaces, in particular in the annular region provided within the reactor vessel between this latter and the reactor core, can be occupied by expanded graphite which is compacted in situ to a predetermined density without preliminary annealing and especially in zones of complex shape, for example around the connecting passages through which the molten fuel salt is circulated. Said passages which connect the reactor core to the heat exchangers and to the circulating pumps can be formed in the mass of expanded graphite either by forming spaces having the requisite size at the time of filling of the reactor vessel or by employing tubes of dense graphite around which the graphite packing is then compacted. It should finally be noted that the use of expanded graphite does not introduce any dimensional limitation in the mass of graphite employed and consequently makes it possible to dispense with joints between blocks in those zones in which the neutron flux density would result in unacceptable temperatures in the fuel salt which is trapped in said joints. If necessary, said mass can be arranged so as to permit slight circulation of the fuel salt and thus prevent stagnation of this latter.