METHOD AND REACTOR FOR THE GENERATION OF ENERGY IN THE PROCESS OF A CONTROLLED NUCLEAR FISSION

Abstract: In a known method for the generation of energy in the process of a controlled nuclear fission by means of neutrons, the improvements reside in that it comprises selection and a directed recovery of thermal neutrons into a thermal neutron region of the fissile core, circulation of a fuel composition is performed over two loops one of which passing through the thermal neutron region and a cooling region and the other passing through the fast neutron region and the cooling region, and a material of fuel compositions of both loops is interchanged. The fuel composition is disposed within at least two loops one of which intersecting the thermal neutron region and the other intersecting the fast neutron region of the fissile core, and said loops are interconnected by a device for interchanging a material of both loops, said moderator is made in the form of a hollow shaper of a directed thermal neutron flux, said fissile core is located in a hollow, and said shaper comprises ducts for shaping a directed neutron flux oriented only in the direction of a fuel composition loop intersecting the thermal neutron region of the fissile core.

Description: METHOD AND REACTOR FOR THE GENERATION OF ENERGY IN THE PROCESS OF A CONTROLLED NUCLEAR FISSION FIELD OF THE INVENTION The present invention relates to the field of nuclear physics and, more particularly, to the physics of energy generation in the fission-type reactors. BACKGROUND OF THE INVENTION A method for the generation of energy in the process of a controlled nuclear fission is known wherein natural or enriched uranium is used as a fissile material. The process of fission is carried out by means of thermal neutrons which are formed in the process of fast fission neutron moderation. Natural water, heavy water, or graphite is used as a moderator. Nuclear energy is converted to thermal energy transmitted to a coolant, i.e. water in this case. Of the major drawbacks of the method are alow efficiency of using fuel, a risk of large-scale accidents and a low environmental acceptability of the method due to an unsolved problem of spent fuel utilization. [See, for example, Enrico Fermy, Leo Szilard, US Patent No. 2708656, Neutronic reactor. 1955; I. G. Belousov, A. S. Doronin, D. P. Krasheninnikov, V. S. Sery, "Configuration Features of HTMS of A Manufacturing Purpose", VANiT, AViT Series, 1983, Issue 3(16), pp. 13 - 14]. A method for generating energy is also known [see, for example, V. M. Novikov, I. S. Slesarev, P. N. Alekseev, V. V. Ignatiev, S. A. Subbotin, "Nuclear Reactors of Improved Safety - Analysing Conceptual Developments", Energoatomizdat. Moscow, 1993] wherein which Pu-239 is generated under fast neutronic breeding conditions during operation of a reactor. A core fuel irradiated is discharged and processed. In processing, fuel formed during irradiation is isolated from an available fertile material (U-238), a fissile material (Pu-239), thereupon uranium is separated from fission products followed by re-forming a fuel mixture in its original enriched concentrations. The depth of depletion is determined by an initial amount of fuel enrichment in a working isotope with regard to a periodical restoration in an open fuel cycle. Fuel has an excess criticality. Of the major draw-backs of the method are a low efficiency of using fuel, a risk of large- scale accidents as well as a low environmental acceptability of the method, because a part of radionuclides finds its way into the environment in the process for chemical processing of ir-radiated fuel. There has been also known a method for the generation of energy using a thorium cycle [see, for example, V. M. Novikov, I. S. Slesarev, P. N. Alekseev, V. V. Ignatiev, S. A. Subbotin, "Nuclear Reactors of Improved Safety - Analysing Conceptual Develop-ments", Moscow, Energoatomizdat, 1993], wherein U-233 obtained from Pa-233 which is generated in a fissile core from Th-232 is used as a major fissile nuclide. Using chemical process¬ ing, Pa-233 is isolated from a fuel mixture and allowed to stand outside the reactor until its conversion into uranium recirculated into a fuel mixture occurs. An insufficient efficiency and ecological acceptability of the method constitute its major disadvantages. There has been known a method for generating energy in the process of nuclear fission by means of thermal neutrons [see, for example, I. K. Kikoin, V. A. Dmitrievsky, I. S. Grigoriev, Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958, Vol. 2, p. 232, Moscow, Atomizdat, 1959] comprising the steps of circulating a fissile material in the gaseous phase within and outside a fissile core, moderating and diffusion recoverying neutrons in the fissile core, introducing the fissile material and converting a released energy followed by leading away fission products. The depth of fuel depletion is determined by its original enrichment and by a total neutron flux during a reactor period. The reactor fuel cycle comprising isolating and a continuous leading away fission products therein requires a regular compensation for a depleted fuel composition comprised of the enriched material. The above method has a disadvantage residing in a low efficiency of using fuel due to the fact that the reactor burns a nonstationary mixture of nuclear fuel which needs to be compensated for with the use of an outer fuel cycle; moreover, this detrimentally affects the environment. The closest prior art dealing with methods for generating energy has been described in a method for the generation of energy using a thorium fuel cycle [see, for example, V. M. Novikov, I. S. Slesarev, P. N. Alekseev, V. V. Ignatiev, S. A. Subbotin, "Nuclear Reactors of Improved Safety - Analysing Conceptual Develop-ments", Moscow, Energoatomizdat, 1993; V. M. Novikov, V. V. Ignatiev, V. I. Fedulov, V. N. Cherednikov, "Physics and Techniques of Nuclear Reactors " , Issue 30, Molten-Salt Nuclear Power Plants: Prospects and Problems, Moscow, Energoatom-izdat, 1990] which is implemented just in the reactor in the process of nuclear fission by means of thermal neutrons, comprising the steps of circulating a fuel composition in the form of salt melts within and outside a fissile core, moderating and diffusion recoverying neutrons in the fissile core. In this case, neutrons, of course, will diffuse into the fissile core being dispersed by a moderating material and reflected in the walls of a reflector. Said method comprises an additional feeding a fissile material, converting a released energy and leading away fission products. Approximately a half of the fissile material is present outside the fissile core, while 13 % of the fuel composition material is present within the fissile core and 37 % within a breeder core. Fast neutron fluxes of thermal and fast regions of the fissile core are inter-crossed. Here, the formation of the fissile core with the fast and thermal regions due to a fissile material configuration results in that the neutron spectrum thus formed is insufficiently mild in the thermal region and insufficiently stringent in the fast region of the reactor. In accordance with this method, a total processing time of the fissile material with the separation of protactinium in a bypass loop equals 10 days. In this case, processing permits, while "cooling" a separated protactinium in its delay line for more than 90 days, an increase in a fraction of its conversion into U-233. The reactor operates under breeding conditions; however, its operation under equilibrium fuel cycle conditions is also provided for. The drawbacks of said method are as follows: the presence of chemical processing during an outer fuel cycle, which detrimentally affects ecology and brings down safety of the process, to say nothing of a low efficiency of using neutrons. There has been known a reactor [Enrico Fermy, Leo Szilard, US Patent No. 2708656, Neutronic reactor, 1955; I. G. Belousov, A. S. Doronin, D. P. Krasheninnikov. V. S. Sery, "Configuration Features of HTMS of A Manufacturing Purpose". VANiT, AViT Series. 1983, Issue 3(16), pp. 13 - 14] for the generation of energy in the process of a controlled nuclear fission comprising a moderator, a fissile core, a fuel composition containing a fissile material in solid fuel elements such that the volume ratio of said fissile material and moderator along with their dimensions and purity permit the occurence of the reaction of fission, and a coolant in a closed loop which is contained in an energy conversion device and intersects said fissile core. The drawbacks of this reactor reside in its low power efficiency, a possibility of large radiation disasters as well as a detrimental influence on the environment. In this reactor, a fissile material is compensated for by means of a complete replacement of assemblies of fuel elements over long periods between which a continuous fuel-nuclide depletion and fission-product buildup take place, which permits limitation of the degree of fuel depletion and calls for reprocessing and reenriching thereof in an outer fuel cycle providing for an additional amount of radioactive waste. In connection with a protracted excess criticality of a fissile material, a substantial fraction of neutrons finds no use in the fission process and is eliminated on the absorbers. Characteristically, control is effected by absorbing excess neutrons in the reactor fissile core, which promotes a risk of possible radiation accidents due to a poor neutron controllability with their purely diffusion recovery in the fissile core. The closest prior art has been described in a reactor for the generation of energy in the process of a controlled nuclear fission [see, for example, V. M. Novikov, I. S. Slesarev, P. N. Alekseev, V. V. Ignatiev, S. A. Subbotin. "Nuclear Reactors of Improved Safety - Analysing Conceptual Developments". Moscow. Energo-atomizdat. 1993; V. M. Novikov, V. V. Ignatiev. V. I. Fedulov, V. N. Chercdnikov, "Physics and Techniques of Nuclear Reactors", Issue 30, Molten-Salt Nuclear Power Plants: Prospects and Problems, Energoatomizdat, Moscow, 1990] comprising a moderator, a fissile core with thermal and fast neutron regions, wherein a distance between said regions is smaller than the lengh of the fast neutron moderation therebetween, a circulation loop with a fissile material which intersects the fissile core such that the volume ratio of said fissile material and moderator along with their dimensions and purity provides for the occurence of the reaction of fission, and a fertile material feeder. The drawbacks of this reactor reside in its low power efficiency, a pronounced risk of accidents and a detrimental influence on the environment. The presence of chemical fuel processing and a high density of fuel material as compared with that of neutron flux significantly impair ecology and safety of the process. Summing up the aforementioned, the following negative trends can be emphasized: 1) a need for enriching an original fissile material and its processing outside the reactor after or in the course of a depletion period in the reactor brings down a power efficiency and environmental acceptability of the process; 2) elimination of a substantial neutron fraction on the absorbers reduces power capacity of the reactor; 3) an excess criticality and high density of a fissile material as compared with that of neutron flux result in an increased radiation hazard. Taken together, these factors lead to the impossibility of full and safe use of nuclear fuel and create both a problem of safety improvement and an environmental problem of radioactive waste utilization. SUMMARY OF THE INVENTION The present invention is made in order to solve the above problems included in the prior arts, and a principle object of this invention is to provide a method and apparatus for generating energy in the process of a controlled nuclear fission with the help of neutrons, which enables to enhance power efficiency, safety and environmental acceptability. A technical result obtainable when implementing the present invention resides in the following: 1) the formation of a critical stationary composition of fuel from an original natural fertile material; 2) the establishment of two separate fluxes of major and breeder material of a fuel composition: 3) the establishment of two separate fluxes of thermal and fast neutrons: 4) the formation of an optimum fission neutron spectrum in a fissile core. The above technical result is achieved in accordance with the present invention by the fact that in a known method for the generation of energy in the process of a controlled nuclear fission by means of neutrons, comprising moderating and recoverying thermal neutrons to a fissile core, circulating a fuel composition containing a fertile material with isotopes of so formed fissile materials through a fast neutron region, a thermal neutron region of the fissile core, and through a cooling region, the improvements reside in that it further comprises selection and a directed recovery of thermal neutrons into a thermal neutron region of the fissile core, circulation of a fuel composition is performed over two loops one of which passing through the thermal neutron region and a cooling region and the other passing through the fast neutron region and the cooling region, and a material of fuel compositions of both loops is interchanged. A further embodiment of the present method is possible in which the ratio of fuel residence time in the cooling region to fuel residence time in the thermal neutron region in a loop passing through the thermal neutron region is maintained beyond 100. A long-term irradiation of the fertile material (natural uranium or thorium) by means of neutron flux under the above- mentioned conditions affords production of a stationary mixture of fissile materials, i.e. a fuel composition having its criticality above 1. A stationary fuel composition contains all isotopes of a substance which were formed after irradiation of a fertile material and its capture of neutrons in the reactor fissile core, and there-after were exposed to gamma-, beta-, alpha-decays of nuclei in sequence. A composition of the mixture so obtained becomes strationary when the rate of formation of each nuclide in the mixture is equal to the rate of its depletion or conversion, and hereinafter said mixture does not practically underdo a time-dependent change of its composition. Irradiation of the fuel composition by fast neutrons makes it possible to generate there neutron-excess nuclei: the exchange of compositions permits transfer of these nuclei to a coolant circuit. The presence of the mixture in a cooling region will convert, thanks to beta-decays, these nuclei to isotopes fissionable on thermal neutrons which burn out in a thermal region of the fissile core, and create a fast neutron flux for a repeated production of neutron-excess nuclei in a fast neutron region from a material of the composition. The increase in a residence time of the mixture outside the fissile core results in the formation from neutron-excess actinides of those actinides which possess a greater charge due to the chain of nuclear beta- decays occured in the mixture and, hence, an increased value of the fission parameter (Z 2 /A) and accordingly a greater fission cross-sections. A parameter with which a stationary composition is controlled is the ratio of a fuel residence time in the cooling region to a fuel residence time in the thermal neutron region. Under the periodic duty, a fraction of actinides with an increased value of this parameter grows. Such a fact takes place in the range of from the time ratios of about 10/1 - 100/1 and grows with an increase in said time ratios. With the time ratios of about 500/1 - 1000/1, a mixture becomes critical over a broad range of densities of the material in said thermal and fast loops. In the latter, there is some decrease in the growth rate of mixture criticality in the fuel composition because a fraction of heavy actinides exposed to alpha-decays becomes significant, which leads to the loss of neutrons out of the mixture. According to the present invention, it is not an absolute residence time of fuel within the fissile core and that outside of this core essential, but the ratio of these times which is directly involved in an equation of isotope generation process. A decrease in the ratio of fast neutron flux quantity to thermal neutron flux quantity in the loop passing the fast neutron region up to the value of above 100 impairs the process for the formation of a fuel mixture due to depletion of actinides in the fast loop until their beta-decays in the thermal loop and due to the growth of a fraction of alpha-radioactive nuclei. The formation of compositions with a criticality of more than 1 takes place in more than 10 8 seconds with neutron flux amounting to about 10 16 neutron/cm 2 per second, depending upon the conditions of its formation and the nature of an original fertile material. The formation of a fuel composition being stationary in the full sense of this notion takes place in more than 3*10 10 seconds, with the interchange of compositions of above 10 7 1 /second. By adjusting a density of the thermal neutron flux, a composition of the fertile material, and a feed into a stationary mixture, by changing the ratio of irradiation time thereof in the fissile core and residence time thereof outside said fissile core in the loops, a degree of material interchange in the loops and a fraction of thermal neutrons in the fast loop, one may control criticality of a stationary fuel composition and maintain steady concentrations of its elements as much as is desired. Depending upon operating conditions used, the process is subdivided into the following stages: 1) an initial depletion of original fissile isotopes of the fuel; said stage is characterized by an excess of neutrons and a positive criticality of the mixture depending on its primary enrichment degree; 2) saturation and stabilization of the composition with intermediate nuclides; said stage is characterized by neutron absorption and a drop of criticality of compositions (the stages may overlap, and a total criticality of the mixture may be > 1 for all the time until the composition is stabilized); 3) depletion of stationary compositions when a fertile material is continuously introduced; said stage is characterized by an excess of neutrons, a controlled positive criticality as well as stability of the process; 4) depletion of the stationary composition of the fuel mixture when the reactor is put out of operation. To reduce a period of forming a stationary composition, a critical mixture of isotopes of fissile materials made of actinides of the stationary mixture of fissile materials may be used as an original composition. Here, one may form a primary mixture comprising actinides separated from an irradiated nuclear fuel. Thus, in accordance with a preferred method, a fuel composition is formed and then maintained stationary thanks to its trans¬ mutation with neutrons and burning out of heavy actinides; this makes it possible to create conditions for a deep depletion of the fuel. Upon loading a reactor and forming a stationary mixture of fissile isotopes followed by controlling the ratio of a residence time of the mixture within and outside a reactor fissile core and the interchange of materials between two closed loops, good grounds for a deep fuel depletion are created. A fertile material may, as it burns out, be introduced, as needed, in both separate loops at once, or one loop after another, while interchanging compositions of both loops. An essential distinction of the present invention from the prior art methods of the same purpose resides in that the present method needs no fuel processing neither for the purpose of introducing said fuel into a reactor nor for a long-term storage or burial of unburnt fuel remainings. Subsequent coolin of fission products and transmutation of long-lived radioactive fission products in accordance with the present invention will convert a major fraction of radioactive fission fragments to stable nuclides. The use of the gaseous fuel composition in the thermal region of the reactor with a low enrichment quantity makes it possible to lower a maximum energy liberation in emergency operating conditions of the reactor, a risk and consequences of possible accidents. This results in the improvement of safety of the entire reactor. A high flow rate of the fissile material in the thermal neutron region of the fissile core allows decrease in a total mass quantity of the fuel composition in the reaction zone. Criticality of the composition varies constantly during its circulation. When the composition moves through the fissile core, its criticality goes down due to depletion of fissile isotopes, and the composition is saturated with fission fragments, thus the composition may be subcritical at the fissile core outlet. When the composition is outside the fissile core, its criticality grows due to beta-decays in a stationary mixture and the removal of fission fragments. Another embodiment of the method is possible, in which a fuel composition is optimized in such a way that criticality of the composition is close to one. Criticality of the mixture in various loops may differ in respect to a loop present in the flux of thermal neutrons, it may be close to or even less than one, while in respect to a loop in the fast neutron region, it may be greater than one, but on fast neutrons. Then, to operate a reactor, fast neutron flux from a thermal loop is necessary, this is provided for by intercrossing fast neutron fluxes of the reactor. Control of the reactor is effected thanks to controlling thermal neutrons. Thereupon, fast neutron fluxes are moderated and directed to a thermal neutron region. Operation of the reactor and the time of forming a stationary composition significantly depend on the density of a flux of neutrons and the efficiency of their recovery in the reactor. A decrease in a fraction of neutrons due to their absorption by fission fragments (up to several tens of percents) is relatively small, therefore the separation and removal of fission products out of the reactor are optional to implement the process, but desirable to reduce neutron losses and enhance power efficiency of the process. The possibility of introducing a coolant into a fuel composition and diluting a fissile material with said coolant as well as a flow-type character of the presence of a material in the fissile core are important to carry out the present method and optimize power removal and compounding. With elevation of temperature of gas, its velocity is rising, the density of a material and, hence, a relative density of a fissile material in the mixture is falling. This permits a certain decrease in a fraction of fuel composition nuclei interacting with neutrons as well as stabilization of a process of their depletion. The negative reactivity temperature coefficient is characteristic of gas- and liquid-fuel reactors. To strengthen the role of this effect, it is possible to further use a pronounced growth of an amount of material in phase transitions which take place with elevation of temperature, e.g., in the dissociation of a coolant material. These and other objects, features and advantages of the present invention will become more clear from the following description of the preferred embodiments taken in conjunction with the accompanying drawings. BRIEF DESCRIPTION OF THE DRAWINGS FIG. 1 is a flow diagram showing the proposed method with the indication of circulation of materials and neutrons; FIGS. 2 and 3 show schematically a nuclear reactor. DESCRIPTION OF THE PREFERRED EMBODIMENTS OF THE INVENTION A nuclear reactor comprises: a directed thermal neutron flux shaper 1 , a focal region 2 of said directed thermal neutron flux shaper, a fissile core 3, a thermal neutron region 4, a fast neutron region 5, a fuel composition 6, a loop 7 with said fuel composition 6 intersecting said fast neutron region 5. a loop 8 with said fuel composition 6 intersecting said thermal neutron region 4, a means 9 for interchanging a material of said loops 7, 8, a means 10 for introducing said fuel composition 6 into said fissile core 3, an energy converter 11 , a means 12 for selecting and leading away fission products, a means 13 for introducing a fertile material 14, a fuel composition storage space 1 , a thermal neutron absorber 16. In the reactor, inside its protective vessel, a moderator is disposed which is made in the form of a directed thermal neutron flux shaper 1 (a moderating-focusing structure, MFS) with a focal region 2 and a fissile core inside said shaper. In the fissile core 3, there are disposed loops 7, 8 containing a fuel composition 6 of a fissile material with actinides in their stationary concentration together with a coolant. The loop 7 with the fuel composition 6 intersects the fissile core in a fast neutron region S, whereas the loop 8 with the fuel composition 6 intersects the fissile core in a thermal neutron region 4. Said loops 7, 8 each contains a means 10 for introducing the fuel composition 6, said means is disposed at the inlet of the fissile core 3, an energy converter means 11, a means 12 for selecting and leading away fission products, a means 13 for introducing a fertile material 14, a fuel composition storage space IS for storing the fuel composition in a cooling region, said loops are interconnected with the help of a means 9 for inter¬ changing a material of the loops 7, 8. The loop 8 with the fuel composition 6 intersecting the fissile core in the fast neutron region 4 may be placed in a thermal neutron absorber 16. The directed thermal neutron flux shaper 1 may also comprise an additional focal region and be placed in the external magnetic field. BEST MODE TO CARRY OUT THE INVENTION The nuclear reactor proposed to implement the present method operates as follows. The fuel composition 6 present in the working loops 7, 8 circulates through the fissile core 3 of the reactor. Originated in the reactor, fast neutrons having passed a material of the fuel composition 6 and partially interacted therewith, enter the directed neutron flux shaper 1, move in the material of the anisotropic moderator and give up its energy taken away from the structure by a coolant flow. Having lossed their energy, thermal neutrons diffuse in a moderating material and being reflected from the surfaces of its anisotropic structure are shaped into neutron flux directed to the focal region 2, which allows to substantially increase the density of neutrons in the abovesaid direction. Fast neutron fluxes thanks to disposition of the loops within the line-of-sight range and due to their reflection from an inrier wall of the directed neutron flux shaper 1 profiled in the form of ellipsoid are intercrossed on a material of the loops. Neutrons which leave the fissile core are then moderated and selected by the directed neutron flux shaper 1. This enables to form a joint flux of thermal neutrons directed to the thermal neutron region. Originated in the fissile core 3, neutrons are thermalized on a moderating material in said directed neutron flux shaper 1 and returned to the fissile core, the focal region 2 where fast neutrons originate again. Thus, a neutron life cycle is repeated. In the proposed reactor, a separate formation of fast and thermal neutron fluxes is realized. Those neutrons returnable in the thermal region, in accordance with a method of forming thereof, contain no neutrons of an intermediate energy, but rather, these neutrons are purely thermal, however the spectrum of intercrossed fast neutron fluxes is hard to the utmost because of the absence of a moderating material in the fissile core. The fuel composition 6 may be introduced into the fissile core in the gas, liquid, or solid phase. The coolant may additionally be introduced from the fissile core periphery. When entered in the gas phase, the fuel composition 6 may be present in the form of volatile compounds, for example fluorides, or in the form of vapors. In this case, helium, for example, may be used as a coolant. When entered in the liquid phase, the fuel composition 6 may be present in the form of low-melting compounds such as, for example, molten-salt compounds. In this case, the coolant may also be molten-saline. The fuel composition introduced in the form of solids may be comprised of pelletized fuel elements, or may, being in the liquid or gas phase, fill up pelletized fuel- element blankets. In this case, helium, for example, is possible as an external coolant. The fuel composition 6 in the loop 7 which passes through the fast neutron region 5 exists mostly in the fissile core 1 and accumulates neutron- excess nuclei formed. To increase interaction in this loop, the density of a material and its cross dimensions must be greater than that in the loop 8 with the fuel composition 6 intersecting the thermal neutron region 4. Gas should be present under a high pressure. A means 10 for introducing the fuel composition 6 may be made in the form of, for example, a jet pump, or a shooting device (in the case of elements and blankets), said means delivers fuel to the fissile core and controls the rate of its introduction thereto. A material of the loop 7 is partially led away with the help of a means 9 for interchanging a material of the loops to the loop 8 with the fuel composition 6. A material from the loop 8 steadily enters the loop 7. In case where the fuel compositions 6 of the loops 7, 8 are in the same phases said means 9 is a typical means for their interdelivery and interchange. However, in case where the fuel compositions 6 of the loops 7, 8 are in different phases, both a plant and a chemical reactor may be used depending on different chemical conditions. The most optimal operation is ensured by the use of a plant which is based on a gas-phase fuel composition. Operating a liquid-phase fuel composition brings down the negative reactivity temperature coefficient, which affects safety; operating compositions having different phases in different loops requires a processing, including chemical processing, in the fuel cycle, which detrimentally affects ecology and brings down safety. A choice of designing a means 13 for introducing a fertile material 14 is determined by the type of the fuel mixture in the loops 7, 8, by its composition and phase, and it may be realized in the form of, for example, a pump for liquid or gas phases containing a pipe connection with the loops 7, 8, or with any one of those loops. It may be optimal to combine it with the means 9 for interchanging a material of the loops by adding thereto an additional inlet for introducing a fertile material in a required phase, or an inlet for introducing a solid fuel mixture. The fuel composition 6 is present in the fissile core 3 during a period of passing (flowing) therethrough. At that time, the temperature attainable in the mixture is determined by energy liberation, heat capacity of the fuel mixture pumpable together with the coolant through the fissile core as well as its residence time within the fissile core. Thanks to the profiling a loop passage in the fissile core, various gas- flow conditions such as, for example, isobaric or isothermic may be realized with respect to gases. Since the residence time of a particular material is only determined by its travel speed through the fissile core and may be as low as 10-' - 10 3 seconds, operation of the reactor of high power is possible under high neutron fluxes and at a relatively low temperature of material at the reactor outlet. A coolant may be introduced additionally, from the fissile core periphery. The fuel composition 6 in the fast neutron region is also present during the time of passing therethrough, but due to the lower interaction cross-sections with fast neutrons, a travel speed of the composition must be sharply reduced, while its density must be sharply elevated. In such a case, the temperature attainable in the mixture is determined by energy liberation, heat capacity of the fuel mixture pumpable together with the coolant through the fissile core as well as by its residence time within the fast neutron region 7, but in connection with a relatively insignificant amount of fast-neutron cross-sections this temperature may be lower than that in the thermal neutron region 8. The residence time of the mixture outside the fissile core must be low, less than that within the fissile core. The fissile core 3 of the loop 8 with the fuel composition 6 intersecting the thermal neutron region 4 makes up only a minor part of the entire loop space, and short-lived beta-radioactive actinides manage to transform, outside the fissile core, i. e. in the region of cooling the fuel composition, into fissile radionuclides, and short-lived fission fragments, in the course of their full circulation, into long-lived, or stable elements. In this case, since the circulation process is durable and continuous, also long-lived actinides do eventually transform into radionuclides fissionable either on fast or slow neutrons. This cooling region may be made in the form of, for example, a space 15 for storing the fuel composition in the cooling region included in the loops 7, 8. When leaving the fissile core 3 in the loops 7, 8, the fuel composition finds its way into an energy converter 11, then into a means 12 for selecting and leading away fission products; after passing through a means 13 for introducing a fertile material 14, a means 9 for interchanging a material of the loops, it is again set out by a means 10 into the fissile core 3, or into the thermal neutron region 5, or into the fast neutron region 4. Thus, the fuel composition life cycle is repeated. Both the entire flow of a working reactor loop and a part thereof may be led away to the means 12 for selecting and leading away fission products. Since fission fragments are significantly distinguished from actinides by their mass, this permits a continuous extraction thereof from a matter of the fuel composition, to separate them from fissile materials therein, and to maintain their quantity at a level which practically has no effect on neutron absorption. From a matter of the fuel composition made in the form of individual fuel elements or blankets, fragments may not be recovered. Extraction of a coolant from the fuel composition followed by the formation of a mixture in novel ratios is also possible. A part of fission fragments, in particular Cs-137 and Sr-90, may be returned to the focal region 2, or to an additional focal region and transmutated there into stable substances. In this case, due to a somewhat longer residence time of a stationary mixture of fissile materials within the means 12 for selecting and leading away fission products, this mixture possesses a higher criticality which makes it possible, while controlling a process of mixing with a major flux of the loops 7, 8 along with the means 10 for introducing the fuel composition determining a travel speed of the fuel composition in the fissile core 3 and of the space 15 for storing the fuel composition in the cooling region, to efficiently control parameters of the reactor. The change of a fissile material residence time within and outside the fissile core 1 admits different embodiments. An energy converter 11 may be any suitable device, for example any heat engine, and may appear directly in the loop, or may be connected therewith through a heat exchanger. In another embodiment said energy converter 11 is characterized in that it comprises a means for the plasma confinement inside the fissile core which is made in the form of a magnetic trap. This permits isolation of a hot plasma region formed under the reactor stringent conditions from the reactor design elements. In this case, a further embodiment is possible in which said plasma confinement means is made in the form of an open magnetic trap a part of which has the form of a magnetic nozzle for converting energy formed into its motion, and is also composed of MHD- or EHD-generator, and/or of some other energy converter. When using the reactor as a rocket propulsor, a part of the moving plasma or of the coolant may be removed into space. The loop 8 with the fuel composition 6 intersecting the fissile core in the thermal neutron region 4 may be placed in the thermal neutron absorber 16 which may be made in the form of a two-layer tube 17 having a liquid or gas neutron absorber 18 connected to a control device 19 of the absorber 18. This permits a further decrease in a fraction of the flux of those thermal neutrons which ingress into the fast neutron loop and thus impair the process of forming a fuel composition. Thermal neutron flux reaching the fast neutron region must be at least one-hundred as many as fast neutron flux reaching the loop. In this case, fast neutron flux entering and leaving the loop may not be substantially moderated by such an absorber. The absorber may be disposed in a hollow tube, pumped therethrough by the control device 19 of the absorber 18, followed by the removal of energy liberated there as heat. The most dangerous emergency situation occurs when there is an abrupt interruption of selecting a meterial from the fissile core 3 and a sharp elevation of the material concentration in this core, therefore designs rutting out the possibility of this scenario are necessary and possible such as, for example, the presence of a bypass loop at the fissile core outlet, concentration of all composite elements at the fissile core inlet, and also both an increased inlet pressure and a reduced exit pressure of a material used, etc. More important, however, is the fact that criticality of a stationary mixture of the thermal loop may, as pointed out above, be maintained at a level close to its minimal values for the given type of reactors, and be self- stabilized by a gas-gas distribution within the mixture and by changing the density of a working mixture composition when heating, which is enhanced by dissociating a coolant substance in the reactor. Furthermore, safety is significantly improved by railing out a fuel enrichment, an external fuel processing, by a considerable simplification of the entire fuel cycle of energy generation in nuclear fission, which also influences ecology of the present method. A process for controlling the reactor may be effected by controlling the ratio of a stationaty composition residence time within and outside the fissile core by means of varying the rate of the fuel mixture passage within and outside said fissile core; by controlling the interchange of a material in the loops; by controlling the choice of composition and characteristics of the coolant, including its working pressure, a rate of pumping through the fissile core, and its dissociation and evaporation temperature, in the latter case a neutron flux self-stabilization due to the negative reactivity temperature coefficient; by controlling the change of a duct cross-section in the fissile core; by controlling a fission-type reactor when using a fertile material as an absorbent introduced while forming its stationary mixture; by controlling processes for selecting in a moderating-focusing structure, MFS; by controlling the quantity of the external magnetic field (or the change thereof) in the embodiment using said MFS with magnetic neutron-reflecting supermirrors; by controlling absorption of thermal neutrons directed to the reactor fast loop; and, finally, by the expense of a typical absorption of excess, from the point of view of fission process, neutrons in the absorber under emergency operating conditions of the reactor. Neutrons may also be absorbed in additional focal regions of the directed thermal neutron flux shaper 1. The advantages of the proposed method and apparatus for generating energy in the process of a controlled nuclear fission consist in the fact that both said method and apparatus permit simplification a fuel cycle of the reactor and an increase in the efficiency of using a fertile material, operation with unenriched fissile material and without a subsequent fuel processing, improvements in an environmental acceptability of the process for generating nuclear energy, elevation of the efficiency of beneficial use of neutrons, as well as improvements in safety of nuclear power-plants. Exam le For a better understanding of the operating principle, some calculations of the processes for irradiating a fuel composition in the reactor having the following parameters were performed. The duct radius in the major (thermal) region of the reactor was 20 cm. The duct length in the major (thermal) region of the reactor was 100 cm. Said duct radius and length in the fast region also equaled 20 cm and 100 cm, respectively. A thermal neutron flux was stabilized at the level of 10 16 neutrons*sq. cm per second; a fast neutron flux was formed depending on the thermal neutrons, but it would not exceed 10 15 neutτons*sq. cm per second. The efficiency of returning neutrons to the fissile core by means of a moderating-focusing structure was taken as 0.8. Natural uranium was chosen as an original fuel and a fertile material. The process was given in the following parametric representation. The density of a fuel composition material in the fissile core of a thermal focus was equal to 1.0* 10 20 atoms/cub. cm. The density of said fuel composition material in a blanket region was equal to 2.0*10 21 atoms/cub. cm. A volume ratio of the fissile core to the thermal loop cooling region was 1: 1000, a volume ratio of the fissile core to the fast loop cooling region was 10 : 1. A fraction of the material interchanged between the above-said two loops amounted to 1*10 7 of a quantity of the material contained in the thermal loop. The rate of pumping the fuel composition through the focal region was equal to 1.0 meters per second. Helium was used as a coolant. Helium pressure was 15 atm. The rate of pumping the coolant was equal to 100 meters per second. A computational modelling of the processes for forming a fuel stationary composition in the present method was carried out. In the modelling, all reactions of transmutating isotopes of fissile materials participating in the process of their interaction with neutrons, from Th-228 to Es-253, were taken into account. Besides, the processes of n-gamma- capture of fission neutrons and alpha-, beta-decays of nuclei in their interaction with thermal (0.025 eV) and fast (0.9 eV) fission neutrons, as well as inter-transitions between nuclei participating in the process were also taken into consideration (n-2n processes were taken into account only for Th-232, U-235, U-238, Pu-239; more complex processes were ignored). The model allowed for various residence times of a fissile material in the fissile core and the cooling region, as well as a ratio of these values. Simultaneously, processes in the fast and the thermal loops were studied under different conditions according to fluxes of acting neutrons, along with a mutual overlapping of fast neutron fluxes, a thermal neutron recovery to the thermal loop region and an interchange of the material between those loops. The model also allowed for energy liberation in the fissile core and a coolant heating-up in conjunction with a fissile material along a duct axis by assuming a value of a process polytropic curve. Calculations showed that under given parameters from the moment of starting the reactor to the moment of reaching a stationary duty, the neutron multiplication factor is always in excess of 1 and is steadily increased reaching an approximately constant value when run into an integrated neutron flux of 2*10 22 . With a flux value of more than 2*10 22 , fl fuel mixture composition does not practically change and remains stable. An exemplary composition of a stationary fuel mixture formed in the course of a long-term irradiation in a two-loop reactor of a fertile material from natural uranium contains: U234-0.12E-03, U235-0105E-02, U236-0.105E-02, U238-0.97E+00, Np237- 0.41E-02, Pu238-0.54E-03, Pu239-0.15E-01, Pu240-0.35E-02, Pu241-0.36E- 03, Pu242-0.11E-03, Am241-0.21E-03, Am243-0.77E-05, Cm-242-0.34E-05, Cm243-0.11E-06, Cm244-0.57E-06 The present invention may be used for the manufacture of nuclear power-plants utilizing natural uranium, depleted uranium, or thorium as fuel. Small-size reactors may be used in trans-portation means. The proposed method may be employed for the preparation of high-density thermal neutron fluxes.

Claims: CLAIMS 1. A method for the generation of energy in the process of a controlled nuclear fission by means of neutrons, comprising moderating and recoverying thermal neutrons to a fissile core, circulating a fuel composition containing a fertile material with isotopes of so formed fissile materials through a fast neutron region, a thermal neutron region of the fissile core, and through a cooling region, characterized in that it further comprising selecting and a directed recoverying thermal neutrons into a thermal neutron region of the fissile core, circulating a fuel composition is performed over two loops one of which passing through the thermal neutron region and a cooling region and the other passing through the fast neutron region and the cooling region, and a material of fuel compositions of both loops is interchanged. 2. A method according to Claim 1 , characterized in that the ratio of fuel residence time in the cooling region to fuel residence time in the thermal neutron region in a loop passing through the thermal neutron region is maintained beyond 100. 3. A reactor for the generation of energy in the process of a controlled nuclear fission comprising a moderator, a fuel composition in a closed loop passing through the cooling region and the fissile core with the thermal and fast neutron regions, characterized in that the fuel composition is disposed within at least two loops one of which intersecting the thermal neutron region and the other intersecting the fast neutron region of the fissile core, and said loops are interconnected by a device for interchanging a material of both loops, said moderator is made in the form of a hollow shaper of a directed thermal neutron flux, said fissile core is located in a hollow, and said shaper comprises ducts for shaping a directed neutron flux oriented only in the direction of a fuel composition loop intersecting the thermal neutron region of the fissile core. 4. A reactor according to Claim 3, characterized in that a hollow of said directed thermal neutron flux shaper is profiled in the form of an ellipsoidal cylinder with its focuses intercrossing said fuel composition loops, and one of said ellipsoidal cylinder focuses is a focal region of said directed neutron flux shaper. 5. A reactor according to Claim 3, characterized in that a hollow of said directed thermal neutron flux shaper is profiled in the form of a cylinder having a generatrix as a two-pronged star in the form of ellipses, fast neutron regions are placed in the star pronges, a center of said star is a focal region of the directed neutron flux shaper and an intercrossing focus of said ellipses, and focuses of said ellipses intersect the loops with a fuel composition. 6. A reactor according to Claim 3, characterized in that a loop in the fast neutron region is disposed inside a thermal neutron absorber.

Inventors: DROBYSHEVSKY JURY VASILIEVICH, STOLBOV SERGEI NIKOLAEVICH, DROBYSHEVSKY, JURY VASILIEVICH, STOLBOV, SERGEI NIKOLAEVICH
Applicants: DROBYSHEVSKY JURY VASILIEVICH, STOLBOV SERGEI NIKOLAEVICH, DROBYSHEVSKY, JURY VASILIEVICH, STOLBOV, SERGEI NIKOLAEVICH
Application year: 1996
Main classification: G21C1
IPC codes: G21C1