RU2689399C1 - Neutron multiplier - Google Patents

Abstract

FIELD: nuclear physics and equipment.SUBSTANCE: invention relates to neutron multiplier. Device contains a neutron multiplication unit with a neutron supply channel from an external source to the neutron multiplication unit active zone and a neutron flux amplified output unit, both placed in a biological protection system. Neutrons multiplication section surrounded by reflector layer is made in the form of hollow body of revolution formed by side wall, which is part of toroid surface, conical upper and lower walls adjoining top and bottom edges of side wall, which are surfaces of truncated cones with equal generatrices. Small bases of cones face each other, and lower circular wall adjacent to upper edge of lower conic wall, and annular wall adjoining lower edge of upper conic wall, located relative to each other with spatial gap. Part of the neutron multiplication section limited by the side wall and the conical top and bottom walls is the active zone compartment, and the other part of this section is the neutrons output compartment from the core compartment.EFFECT: increasing neutron flux intensity.12 cl, 6 dwg

Description

The invention relates to the field of nuclear physics, in particular to neutron multipliers that can be used in nuclear facilities for various purposes.

 For a number of practical applications, for example, neutron diffraction studies, activation of materials, medical applications of neutron capture therapy, intensive neutron fluxes are necessary. At present, such flows are generated by pulsed neutron tubes, accelerators of various designs, removed from the operating atomic reactor, or obtained on devices that briefly form supercritical assemblies of fissile materials. The described techniques have their drawbacks: neutron tubes have a limited and relatively small life, accelerators and atomic reactors of any type and power are very expensive to manufacture and operate. In this connection, structures providing a constant neutron flux of the required intensity are of practical interest.

 Known technical solution implemented in the construction of neutron multipliers, disclosed in US Pat. RF2178209 C2, publ. 01/10/2002, providing neutron multiplication under subcritical conditions, containing a chamber (10), a fuel-moderating unit (13) with a target (14) in the middle part of the unit, in which a hole is made for the flow of neutrons to the target; A collimator (5) is provided to prevent neutron leakage from the multiplier through the tube (4). The target material is permeable to low-energy neutrons, which are scattered in the opposite direction from the moderator and fuel. The resulting neutron flux is output through the output channel (15). The execution of a neutron multiplier with such a structure is limited by criticality due to relatively small values of k (k <1) and, therefore, by reactivity, due to the provision of only subcritical conditions, which limits the receipt of neutron fluxes with a high level of intensity in the multiplier.

 Receiving neutron fluxes with a high level of intensity can be implemented using subcritical systems based on multi-stage multipliers. From pat. RF2261485 C2, publ. 09/27/2005, a technical solution is known relating to a subcritical multi-stage neutron multiplier containing several adjacent multiplication cascades. Multiplication cascades consist of adjacent coaxially different-thickness discs. The disks are made of active material, two types of neutron moderators and a neutron absorber. They alternate in a cascade in such a way that a unidirectional passage of neutrons and their multiplication is created. The technical result is to provide a large multiplication factor of neutrons. Fast neutrons consistently pass the absorber, two moderators, in which they dump their energy, slow down to thermal energies and approach the active layer. In the active layer, thermal neutrons are absorbed, causing the fission of uranium-235, multiply in it, compensating for all neutron losses due to absorption and scattering along the path from the source to uranium. The absorber is made of boron-10 material. The first moderator can be made of polyethylene or water, the second moderator is made of beryllium. The active layer is made of uranium-235 90% enrichment. The group of the absorber and the two moderators do not allow the reverse movement of thermal neutrons to the previous cascade. The implementation of neutron multipliers according to this scheme allows to solve the problem of increasing the level of intensity of the neuron flux, but leads to an increase in the multiplier dimensions, which limits its scope. In addition, the device contains a large amount of highly enriched uranium (more than a few critical masses in a compact arrangement) and is potentially dangerous when destroyed.

 The present invention is to develop a neutron multiplier that provides a constant high-intensity neutron flux without a significant increase in its size and amount of fuel used.

 The technical result is to increase the intensity of the neutron flux, provided by the presence of supercritical propagation zones on thermal neutrons with a total amount of fissile material less critical mass.

 This technical result is provided by the device neutron multiplier containing

 a neutron multiplication unit placed in the biological protection system with a neutron feed channel from an external source to the core of the neutron multiplication unit, and an enhanced neutron output unit, in which

 the fast neutron multiplication unit includes neutron multiplication section placed in a sealed container and the neutron output section communicating with the neuron output unit, and

 the neutron multiplication section, surrounded by a layer of reflector, is made in the form of a hollow body of revolution, formed

  side wall, which is part of the surface of the torus,

  conical upper and lower walls adjacent to the upper and lower edges of the side wall, which are the surfaces of truncated cones, with equal generators, whose small bases are facing towards each other, and

  the lower circular wall adjacent to the upper edge of the lower conical wall, and the annular wall adjacent to the lower edge of the upper conical wall, located relative to each other with a spatial gap, and

  The part of the neutron multiplication section, bounded by the side wall and the conical upper and lower walls, is the compartment of the core with the fuel solution in the form of a fissile isotope or mixture of fissile isotopes in the moderator, and the part of this section bounded by the annular and circular walls is the neutron exit section from core compartment,

 The neutron output section is located in said neutron exit compartment coaxially with the neutron multiplication section between said lower circular wall and the upper annular wall of the neutron exit compartment bounded by a conical wall, which is the surface of a truncated cone whose circumference of the larger base is the circumference of the opening of the upper annular wall of the neutron exit compartment,

 herewith, the geometric parameters of the mentioned sections are determined on the basis of the average length ln of the neutron thermalization in the corresponding loaded active medium, so that

 the length of the generatrix of the mentioned conic walls with a slope of 50-55 ° is equal to ln

 the length of the active medium compartment along the centerline of the section lying in the equatorial plane of the torus surface of the side wall is greater than ln,

 the diameter of the said lower circular wall is 2.5 ln,

 the diameter of the hole of the upper annular wall is equal to 2.0 ln,

 the height of the spatial gap between the lower circular and upper annular walls is 0.5 ln, and the angle of inclination of the generatrix of the conical wall of the neutron hatching compartment is 45 °,

 The enhanced neutron output unit includes a neutron tube connected to the neutron output compartment hole of the neutron multiplication section, and an anisotropic filter that closes the neutron tube at its output section.

The propagating medium of the active zone is a fuel solution in the form of a fissile isotope or a mixture of fissile isotopes in the form of a homogeneous solution in a moderator, and the fissile isotope is selected from the group comprising ²³⁹ Pu and ²³⁵ U, while the total number of fissile isotope in the multiplier is less critical. The most preferred inhibitor / solvent is water.

 To compensate for the loss of reactivity in the neutron multiplication zone by returning part of the neutron flux directed to the enhanced neutron output section, the multiplier further comprises a group of retractable reflectors in sealed cylinders installed in the neutron output compartment between the core compartment and the enhanced neutron output section.

 To ensure the maintenance of the constancy of the composition and temperature of the fuel solution in the areas of multiplication and outflow of neutrons, cooling systems and fume conversion units, as well as an emergency fuel discharge system, can be installed.

 The remaining features of the present invention will be understood from the subsequent detailed description of the invention.

 The invention is illustrated by drawings, in which identical elements and parts of the neutron multiplier are indicated by the same reference numerals.

 FIG. 1 is a schematic view of a neutron multiplier,

 FIG. 2 - frontal section of the reproduction unit (fast) neutrons,

 FIG. 3a - the structure and geometry of the generatrix of the surface of the neutron multiplication section;

 FIG. 3b is a frontal section of the neutron multiplication section, showing the surfaces of the multiplication section;

 FIG. 4a is a schematic sectional view of the neutron multiplication section, which shows the geometrical parameters of the section and the neutron multiplication zone in the core compartment;

 FIG. 4b shows a section along the line A-A of the neutron multiplication section in FIG. 4a;

 FIG. 5 is a schematic sectional view of a breeding unit with a fuel insert in a neutron tube;

 FIG. 6 is a schematic sectional view of the breeding unit with the fuel insert at the output of the output section of the enhanced neutron flux.

 A neutron multiplier device whose circuit is shown in FIG. 1 contains a fast neutron multiplication unit 1 with a channel 7 for supplying neutrons from an external source to the active zone of the multiplication unit and a unit 2 for outputting an amplified flux of fast neutrons that are surrounded by means 6 of biological protection.

 The output section of the enhanced neutron flux includes a neutron tube 4 and an anisotropic filter 5. The multiplication unit of fast neutrons with a neutron tube 4 is placed in a sealed container 3. The flux of fast neutrons is output through a neutron tube 4 through an anisotropic filter 5 to the consumer of the amplified neutron flux.

 The fast neutron multiplication unit 1, whose circuit is shown in FIG. 2, contains a fast neutron multiplication section 8a with a neutron output compartment 8b and an enhanced flow flux section 9 located in the neutron output compartment 8b.

 The neutron multiplication section 8 is made in the form of a hollow rotation body, the walls of which form a surface with a complex (combined) section cross section profile. Essentially, the surface of the neutron multiplication section 7 is the result of the rotation around the vertical axis Op-Op of this section forming the open profile corresponding to the cross-sectional profile of this section to its axis Op-Op. As shown in FIG. 3a forming the surface of the breeding section is formed by an arc 8c of a circle, the ends of which pass into straight-line segments 8c1 and 8c2 of equal length with a slope at an angle α to the midline lp-lp of the generatrix of the profile, which in turn turn into horizontal segments 8c3 and 8c4, respectively, and separated by a gap h at that segment 8c4 passes to the axis of rotation Op-Op and is greater than the length of segment 8c3, the free end of which is removed by a distance lh from the axis of the Op-Op section. In general, as seen in FIG. 3b, showing a frontal section of a neutron multiplication section 8 with an enhanced flow output section 9, a neutron multiplier section 8 contains a side wall PT, which is part of the torus surface with an arc profile 8c and an equatorial diameter of the torus surface De = 2Rp, conical conical walls from the upper and lower sides of the section (with a cone angle β = 90 - α), the small bases of which are facing towards each other, the lower circular wall Pcr, and the upper ring Pc wall, with a hole diameter dh = 2lh or another dhodyaschey flat geometrical figure.

 FIG. 4a shows a section along the line A-A coinciding with the middle line lp-lp of the section lying in the equatorial plane of the wall of the breeding section formed by the torus surface.

The space of section 8, bounded by the side wall of PT, which is the surface of the torus, and the conical upper and lower walls of Pcon forms a compartment 8a for filling with multiplying medium 8f, hereinafter referred to as the core compartment. As a propagating medium, there is a fuel solution, for example, in the form of a fissile isotope or a mixture of fissile isotopes in the form of a homogeneous solution in a moderator, in which an isotope or mixture of isotopes is used as a fissile substance, such as isotopes from group ²³⁹ Pu, ²³⁵ U, etc. This is the total number of fissioning isotope less critical. The core compartment 8a has a length L in the equatorial plane of the torus surface of the side wall. The neutron output compartment 8b from the core compartment 8a provides an output of neutrons to the enhanced neutron output section 9f and is a circular cavity with an opening in the upper annular wall of the PC, which is connected to the neutron tube 4 of the enhanced neutron output section, and a continuous lower circular wall, and divided with a compartment of the active zone circular side wall of metal.

 Compartment 8b is a body of rotation, the cross section of which is perpendicular to the axis of rotation has the shape of a rectangular trapezoid with sides of 0.25 ln (upper base of the trapezoid), 0.75 ln (lower base of the trapezium), 0.5 ln (lateral side of the trapezium adjacent to the bases under the straight angle). The compartment is made of low-activated stainless steel, either zirconium or any other alloy, which has high mechanical strength and a small neutron capture section. The cavity of the compartment is filled with helium at a pressure of from 1 Pa to 100 kPa.

 The upper and lower walls of the neutron exit section are separated by a height equal to the above-mentioned gap h. = 0.5 ln.

 The neutron multiplication section is surrounded by a layer of a reflector 10 made of graphite or other materials providing at least a neutron flux albedo of 0.7. The thickness of the reflector layer does not exceed 50 cm.

 Section 9 output enhanced neutron flux is located in the Central part of the compartment 8b neutron output from the compartment 8a of the active zone between the upper annular wall and the lower circular wall of the neutron output compartment and is made in the form of a hollow compartment 9a bounded by a conical wall, which is the surface of a truncated cone, coaxial breeding section 8, and open from the upper side around the circumference of its larger base with a diameter equal to the diameter dh of the opening of the upper annular wall of the neutron exit channel 8b. The cavity of compartment 9a of neutron outflow section 9 is filled with a fuel solution in the form of a fissile isotope or a mixture of fissile isotopes in the form of a homogeneous solution in moderator 9f. The main purpose of the enhanced neutron output section is to ensure the maximum possible leakage of fast and semi-thermal neutrons.

 The geometry of neutron multiplication section 8 and neutron outflow section 9 are determined based on the average ln path of thermalization of the fast neutron for the selected type of fuel solution.

 The lengths of the 8c1 and 8c2 segments corresponding to the generators of the conical walls of the neutron multiplication compartment 8a are taken equal to the length ln of neutron thermalization, the length L of the core 8a of the neutron multiplication section along the midline lp-lp lying in the equatorial plane of the torus surface of the side wall Fri is greater than ln , while the angle α of inclination of the generatrix of the conic surface of the Pcon to the midline lp-lp is 50-55 °. The height h of the neutron output compartment 8b from the core compartment and the neutron extraction section 9 is assumed to be 0.5 ln. The size of the hole in the upper wall of the neutron output compartment 8b from the active medium compartment, equal to the diameter of the larger base of the conical section of the output compartment 9a of the amplified neutron flux, is equal to twice the average neutron thermalization path length - 2ln, and the width of the annular surface of the upper wall of the neutron output compartment 0,25ln. The angle of inclination of the generatrix of the conical wall of the compartment 9a is assumed to be 45 ° and the diameter of the circumference of the smaller base of the conical wall of the output section 9 of the amplified neutron flux will be ln for the above compartment height h.

 With such geometrical parameters in the compartment 8a of the active zone of the neutron multiplication section, in which there are three multiplication regions: the supercritical neutron multiplication region 8-1, the critical neutron multiplication region 8-2 and the neutron multiplication region 8-3, and the general reactivity of the neutron multiplication region is ensured less than 1. A neutron coming into zone 9f, of any energy, is capable of causing only a single act of fission and reproduction, but not a developed chain reaction. Section 9 of the outflow of the enhanced neutron flux at the specified geometry of the conical compartment 9a ensures maximum leakage of the fast neutrons formed in it in the passive mode.

 The length ln of the mean path of thermalization of a fast neutron depends on the material of the moderator used in combining the fissile material with the moderator. As an example, below are the possible ln values of the average thermalization range and their optimal values, which can be taken into account when calculating the geometry of the neutron multiplication unit, for three types of moderator material used in fuel solutions.

Material
retarder Interval
average run
ln cm Optimal
average mileage
ln cm Water (H ₂ O) 5-15 6 Heavy water (D ₂ O) 8-30 eleven Graphite 15-40 nineteen

 The neutron multiplication unit formed in this way is a supercritical generator of the geometric type, containing at least one supercritical propagation zone on thermal neutrons and containing the total amount of fissile material less than the critical mass.

 The breeding unit uses an external source of neutrons, the flow of which enters through channel 7 of neutron supply from an external source to the core compartment 8a with the fuel solution in the form of fissile material in the moderator, triggering the neutron multiplication cycle.

 In the neutron output compartment 8b between the active medium compartment 8a and the neutron outflow section 9 into the neutron tube 4, the neutron flux interruption inserts 17 are placed. Inserts 17 interrupt neutron flux can be made in the form of retractable reflectors placed in sealed cylinders of weakly activated steel or zirconium. Retractable reflectors are made of graphite, beryllium or their compounds and mixtures. The cylinder can be made with a diameter of 1-5 cm, the depth of immersion / extension is equal to the thickness of the insert interrupt the neutron flux. Retractable reflectors return part of the neutron flux to the area of propagation of the active medium compartment to compensate for the loss of reactivity as the fuel burns out. The number of cylinders can be from 6 to 30 pieces, each equipped with an independent displacement drive. FIG. 4b shows, as an example, inserts 17 for interrupting a neural stream of eight retractable reflectors located along the perimeter of zone 8b. As the reactivity is lost in the neutron multiplication region 8a, the retractable reflectors move out of the sockets and ensure the return of part of the neutron flux directed to the output section of the amplified neutron flux.

 The high-energy radiations that form the unit during operation cause radiolysis of water as they pass through aqueous solutions (fuel isotope compounds), leading to the decomposition of water into hydrogen and oxygen and the appearance of explosive gas. To prevent accumulation of detonating gas, the breeding compartment 8a and the neutron ejection section, 9a, are equipped with catalytic conversion of detonating gas 12 and 14, which catalyses the conversion of detonating gas back to water on platinum-metal catalysts. The catalytic unit 12 is installed in the zone of the upper edge of the torus surface of the side wall of the compartment 8a of the active medium of the breeding section 8, and the catalytic unit 14 is installed on the cover of the compartment 9a. The catalytic unit 12 of the active medium compartment can be made in the form of a ring 1 cm thick and 3 cm high, and the catalytic unit 14 can be made in the form of a ring 1 cm thick and 3 cm high and 30-50 cm in diameter.

 The catalytic unit is filled with activated carbon coated with platinum in a concentration of 0.1% by weight of coal. The edge of the ring of the catalytic unit 12, bordering the edge of the torus surface of the compartment 8a of the active medium, and the edge of the ring of the catalytic unit 14, bordering the surface of the outflow of neutrons of the compartment 9a, are perforated and permeable to water. The detonated gas released during the operation of the plant diffuses into the catalytic unit area, undergoes a reverse transformation there and returns as a water condensate back to the fuel solution in the active medium compartment and the fuel solution in the enhanced neutron output section. The catalytic unit also additionally carries out the sorption and removal from the reaction zone of the poisoning elements of a group of radioactive noble gases.

 To ensure automatic irreversible killing of the installation in case of abnormal overheating, the breeding unit can be equipped with an emergency fuel discharge system.

 The emergency fuel discharge system includes an emergency fuel discharge tube 15 connected to the neutron multiplication section 8a of the active medium and a capacity 16 for the emergency discharge of the fuel solution. Tube 15 emergency dumping of fuel can be performed, for example, from corrosion-resistant weakly activated steel with an inner diameter of 30-50 mm. The tube 15 is connected to the capacity of 16 emergency drain fuel solution, made in the form of a sealed vessel, in which there is a dry boric acid or borax in an amount sufficient for deep subcriticality of the fuel solution under any conditions and external influences.

 The amplified fast neutron flux is removed from the output section of the enhanced neutron flux through the neutron tube 4, which is closed at the output section by an anisotropic filter 5.

 The neutron tube 4 is made of weakly activated steel or zirconium; the tube is evacuated or filled with an inert gas, such as helium, neon or argon, at pressures up to 1 atmosphere. The principle of operation of the neutron tube is that the neutrons entering it do not experience the action of the medium, since the tube is evacuated, and as a result, their free path is determined only by their energy.

 Anisotropic filter 5 is designed to provide one-sided neutrons in the direction of the device and should cut off the possible return flow through the neutron tube. The anisotropic filter is made of a thermal neutron absorber layer and a moderator layer.

 The materials of the moderator layer can be liquid or solid, as well as solutions, disperse systems and composite materials such as, for example, water, heavy water, paraffin and related high-boiling alkanes, polyethylene and related polymers (polypropylene, polyisobutylene), suspensions of polymers in water, solutions of lithium hydroxide, alcohols and polyglycols in water, solutions of polymers in organic solvents. The moderator layer has a thickness sufficient to slow down to thermal velocities (0.025-0.04 eV) of neutrons with an initial energy up to 1 MeV. In the case of using polyethylene, or related alkanes, polyalkanes, and their solutions and mixtures, the required layer thickness is 90-100 mm. It is preferable to use cast polyethylene or dispersion of polyethylene in paraffin with a layer thickness of 5-10 cm.

The materials of the absorber layer (with a large capture section of thermal neutrons) can be liquid or solid, as well as composites. such as, for example, elemental boron, cadmium, rare-earth elements, as well as their compounds, alloys and composites. The absorber layer has a thickness sufficient to absorb a flux of 10 ¹⁴ neutrons per second, i.e. total flow area anisotropic filter. In the case of metallic cadmium is a layer of solid metal with a thickness of 20-25 mm, preferably, a thickness of 10 mm.

 The anisotropic filter is located at the exit of fast neutrons from the neutron tube 4, the absorber layer of which faces the output section 9 of the enhanced neutron flux. The anisotropic filter housing is made of poorly activated structural materials: polymers, weakly activated steel and has sufficient strength to prevent leakage of components of the moderator layers and the absorber. The flux of fast neutrons (2-4 MeV) from the source falls on the absorber layer and passes through it, without losing energy and the integral intensity of the flux. Then it falls on the layer of the moderator and passes it, losing energy, but little losing the overall intensity of the stream, leaving the device. The flux of return neutrons (reflected, generated by the activation of materials behind the layer of the moderator) falls on the layer of the moderator and passes it, slowing down to an energy of 0.025-0.04 eV. In addition, it dissipates in addition, but the overall intensity drops slightly. Then the slowed neutrons are completely absorbed by the absorber layer.

 In the layer of the moderator can be placed channels of active cooling, made in the form of non-volatile heat pipes, with a working fluid of acetone or alcohol. The absorber layer due to its high thermal conductivity, if necessary, is cooled due to the removal of heat beyond the device boundaries by the thermal conductivity of the metal itself.

 An anisotropic filter of neutrons, ensures the attenuation of the flow only when falling from a certain direction. In the case of neutron irradiation from other directions other than the shielding direction, there is no significant attenuation of the flux, although the neutrons are partially scattered and slowed down on the materials of the device.

 The neutron multiplication unit and the enhanced neutron output unit are placed in the biological protection system. Biological protection is mounted on-site, including from local materials and is not an independent neutron multiplier module.

 Biological protection is made of a complex of materials that have the ability to absorb ionizing radiation of various types. The protection is multi-layered; the first layer is made of borated polyethylene, at least 5 cm thick (then 4 to 10 cm) and the content of elemental boron (or its compounds) is at least 4% (mass) (range 3-15%).

  The purpose of the first layer is the complete absorption of neutrons passing through the layer of the reflector. The second layer is made of super heavy concrete, with a density of at least 3.5, and preferably 4.5-5.2; Concrete is prepared by introducing into the composition of cement binder crushed iron ore, cast iron shot and other dense metal or ore materials. The thickness of the layer is at least 20 cm, preferably 40 cm. The range is 20-100 cm. The purpose of the second layer is absorption of (secondary) X-rays and gamma rays.

Neutron multiplier operation

 Through the channel 7 for supplying neutrons, for example in the form of an introductory tube, the neutron flux from an external source is introduced into the core compartment 8a. Neutrons can be thermal, intermediate, or fast, and the flow itself can be both constant and pulsed.

 In the core section of the breeding section, neutrons are thermalized and absorbed by the fuel atoms dissolved in the moderator (water, heavy water, their mixtures and suspensions with graphite). At the same time, second-generation neutrons are generated and there is a general multiplication of the integral neutron flux with a multiplication factor of 2-2.8. The second-generation neutrons are emitted isotropically and, through several fission events, spread throughout the entire neutron multiplication section.

 Since the geometrical dimensions of the core 8a of the reproduction section core and the parameters of the reflector 10 are selected in this way, the multiplication area of the core 8a without external neutron feed from the source is subcritical due to the emission of a significant part of neutrons to the section 9 of the enhanced neutron output.

 In compartment 8a of the multiplication active zone, a supercritical reproduction area, a critical reproduction area, and a subcritical reproduction area are formed. The overall reactivity of the system is less than 1. The region of supercritical reproduction has such a geometrical arrangement and size that half of the neutrons produced in it go into the outflow zone and eventually leave the installation. The remaining part of neutrons passes through the fuel solution, is thermalized and generates a wave of secondary neutrons in the subcritical reproduction area. Neutrons, born in the subcritical reproduction area, are emitted into the supercritical reproduction area, where they thermalize, cause fuel division and close the multiplier operation cycle, and into the reflector. Reflected neutrons are thermalized predominantly in the region of critical reproduction, from which, in turn, are emitted into the region of outflow and into the reflector. Neutrons born in the field of critical reproduction, mostly do not cause fission of the material in the field of supercritical reproduction, since they leave it before reaching thermalization.

 Neutrons trapped in the output compartment of an amplified neutron flux, regardless of their initial energy, are thermalized and cause the fission of the fuel in the emission zone. Neutrons born in the exit section of the enhanced neutron flux leave the installation through a neutron tube 4, closed by an anisotropic filter 5.

 The geometry of the compartment 9a of the output of the enhanced neutron flux and the presence of inserts 17 of interrupting the neutron flux in the neutron exit compartment provide a neutron flux into the active zone of the neuron multiplication compartment smaller than it enters the exit compartment of the enhanced neutron flux. This causes the overall subcriticality of the system, since due to the increased neutron leakage, the reactivity of the system decreases and becomes less than 1. Systems with such reactivity are subcritical and incapable of spontaneous chain reaction.

 In the inserts 17, additional retractable reflectors can be installed, which increase the neutron return to the active zone of the breeding compartment 8a and thereby regulate the degree of its subcriticality as the fuel burns out.

 Pipes 11 and 13 of the cooling system, blocks 12 and 14 of conversion of detonating gas are also an emergency discharge system of fuel 15-16, when they are installed, maintain the constancy of the composition and temperature of the fuel solution in the active zones of multiplication and outflow of the enhanced neutron flux. The emergency fuel discharge system provides automatic irreversible killing of the neutron multiplier in the event of abnormal overheating, which for some reason has arisen in it.

In the framework of the above-described version of the neutron multiplier, the installation begins with the injection of neutrons through a channel 7 for supplying neutrons from an external source that passes into the active zone of the reproduction section 8a compartment and causes the fissionable material to be fissioned, for example ²³⁵ U, ²³⁹ Pu and their mixtures with concentration an element in the range of 1-100 g / l, a preferred range of 1-10 g / l; The optimum range is 3-5 g / l. After that, secondary neutrons are isotropically scattered across the volume of the core of compartment 8a, causing a secondary fission event mainly in the subcritical reproduction area. The number of tertiary neutrons increases in comparison with the secondary ones, however, the total flux density falls. Approximately 55% of tertiary neutrons from the subcritical reproduction area go towards the reflector (and are reflected back with a loss of 30% of intensity) and are emitted approximately 40% towards the supercritical reproduction area. The main power of reproducing and amplifying the neutron flux occurs in the circulation of neutrons between the regions of supercritical reproduction and subcritical reproduction. The location of the channel 7 for supplying neutrons from an external source is not important, since in any case, the introduced beam will come slow in the center of the active zone.

 Neutrons from the supercritical reproduction area (tertiary) cause the fissionable material fission event and the 4th generation of neutrons goes 50% into the compartment 9 of the output of the enhanced neutron flux, where it generates the 5th generation of neutrons leaving the neuron multiplier through the system of the neutron tube 4 and an anisotropic filter five.

 To increase the intensity of the enhanced neutron flux, the neutron multiplier can be equipped with a fuel insert. The fuel insert can be installed in the output compartment of the amplified neutron flux, or at the output of the output compartment of the amplified neutron flux, or in their combination.

 FIG. 5 shows a cross section of a neutron multiplier in which the fuel insert 18 is installed in the output compartment 9a of the amplified neutron flux. The fuel insert 18 is installed vertically along the axis of the compartment 9a and is made of a metal or alloy containing a highly enriched fissile element in amounts comparable to the number of elements in the fuel solution of the outflow zone of the enhanced neutron flux.

As the material of the fuel insert 18 for this embodiment, metallic plutonium can be used, or its compact compounds in the form of: oxides, carbides, alloys with other metals. The dimensions of the fuel insert, made in the form of a cylinder: diameter 1-2 cm, height - 0.5 from the average length ln run of thermalization and depending on the environment of the fuel solution can be selected from the ranges of 2-4 cm, when the material of the solution medium is light water ( H ₂ O), 3-8 cm, when the material of the solution medium is heavy water (D ₂ O), and 8-11 cm, when the material of the solution medium is graphite; Preferably, within each range, the optimum height, respectively, of the aforementioned materials of the solution medium is 3 cm, 5.5 cm, 9.5 cm.

 With this geometry and structure of the neutron multiplier, the 4th generation of neutrons, as compared with the above-mentioned embodiment, goes 50% into the zone of section 9 of the release of an enhanced neutron flux, where it generates the 5th generation of neutrons, in the environment of the fuel solution and the fuel (metal) insert 18 leaving the neutron multiplier through the system of the neutron tube 4 and the anisotropic filter 5.

 In such an embodiment, the neutron yield is increased by 5-20%, and the unit operation time by 1-3 months.

 FIG. 6 shows a cross section of a neutron multiplier in which the fuel insert 19 is installed at the output of the output compartment 9a of the amplified neutron flux. In this embodiment, the fuel insert 19 is parallel and coaxial to the outlet of the output compartment of the enhanced neutron flux and is made of metallic plutonium or its compact compounds: oxides, carbides, alloys with other metals.

The fuel insert 19 is made in the form of a disk with a diameter equal to the average path length ln of the neutron thermalization path and depending on the medium, the fuel solution can be selected from the ranges of 5-7 cm when using the medium of light water solution (H ₂ O) as a material, 9- 13 cm when using a solution of heavy water (D ₂ O) as a medium, and 17-21 cm when using graphite solution as a medium. Preferably, within each range of the mentioned materials of the solution medium, the optimum diameter is 6 cm, 5-11 cm and 19 cm, respectively. The thickness of the disk can be taken in the range of 1-20 mm, preferably 2-3 mm. The distance from the fuel insert to the surface of the fuel solution of the leak zone ranges from 0.5 to the average length ln of the neutron thermalization path in the fuel solution medium, and depending on the moderator material it can be selected from 2-4 cm ranges when the material of the solution medium is light water (H ₂ O), 3-6 cm, when the material of the medium of the solution is heavy water (D ₂ O), and 7-11, when the material of the medium of the solution is graphite; preferably within each range of the solution material mentioned by the medium, the optimum distance is 3 cm, 5.5 cm and 9.5 cm, respectively.

 In this embodiment, compared with the previous version, the neutrons from the supercritical reproduction area (tertiary) cause the fissionable material fission event and the 4th generation of neutrons goes 50% into the outflow zone 9, where it generates the 5th generation of neutrons, in the environment of the fuel solution and the 6th generation of neutrons on the fuel (metal) insert 19, leaving the installation through the system of the neutron tube 4 and an anisotropic filter 5.

 In this embodiment, the neutron yield is increased by 20-40% with a relatively small increase in the operating time of the neutron multiplier.

 The neutron multiplier described above and its individual versions do not limit the present invention. For specialists in this field of technology can be understood possible modifications, changes and improvements that do not go beyond the scope of the legal protection of the present invention in accordance with the claims.

Claims (27)

1. A neutron multiplier containing a neutron multiplication unit placed in a biological protection system with a neutron feed channel from an external source to the core of a neutron multiplication unit and an enhanced neutron output unit, in which the fast neutron multiplication unit includes neutron multiplication section placed in a sealed container and the neutron output section communicating with the neuron output unit, and the neutron multiplication section, surrounded by a layer of reflector, is made in the form of a hollow body of revolution, formed side wall, which is part of the surface of the torus, conical upper and lower walls adjacent to the upper and lower edges of the side wall, which are the surfaces of truncated cones, with equal generators, whose small bases are facing towards each other, and the lower circular wall adjacent to the upper edge of the lower conical wall, and the annular wall adjacent to the lower edge of the upper conical wall, located relative to each other with a spatial gap, and The part of the neutron multiplication section, bounded by the side wall and the conical upper and lower walls, is the compartment of the active zone, and the part of this section, bounded by the annular and circular walls, is the neutron exit compartment from the core compartment, The neutron output section is located in said neutron exit compartment coaxially with the neutron multiplication section between said lower circular wall and the upper annular wall of the neutron exit compartment bounded by a conical wall, which is the surface of a truncated cone whose circumference of the larger base is the circumference of the opening of the upper annular wall of the neutron exit compartment, the geometrical parameters of the mentioned sections are determined on the basis of the average length ln of the thermalization of the neutron in the corresponding fuel solution of the core so that the length of the generatrix of the mentioned conic walls with a slope of 50-55 ° is equal to ln, the length of the active medium compartment along the centerline of the section lying in the equatorial plane of the torus surface of the side wall is greater than ln, the diameter of the said lower circular wall is 2.5 ln, the diameter of the hole of the upper annular wall is equal to 2.0 ln, the height of the spatial gap between the lower circular and upper annular walls is 0.5 ln, and the angle of inclination of the generatrix of the conical wall of the neutron hatching compartment is 45 °, The enhanced neutron output unit includes a neutron tube connected to the neutron output compartment hole of the neutron multiplication section, and an anisotropic filter that closes the neutron tube at its output section. 2. The multiplier according to claim 1, wherein the anisotropic filter is double-layer, the absorber layer and the thermal neutron moderator layer, the absorber layer being made 20-25 mm thick from a material selected from the group including elemental boron, cadmium, rare-earth elements, as well as their compounds, alloys and composites, and the retarder layer is made of 90-100 mm thick from a material selected from the group that includes water, heavy water, paraffin and related high-boiling alkanes, polyethylene and related polymers, suspensions of polymers in water, solutions of lithium hydroxide, alcohols and polyglycols in water, solutions of polymers in organic solvents. 3. The multiplier according to claim 1, further comprising neutron flux interruption inserts in the form of retractable reflectors in sealed cylinders installed in the neutron output compartment between the core compartment and the output section of the amplified neutron flux into the neutron tube. 4. The multiplier according to claim 3, in which the retractable reflectors are made of a material selected from graphite, beryllium or their compounds and mixtures. 5. The multiplier according to claim 1, in which the propagating medium of the active zone is a fuel solution in the form of a fissile isotope or a mixture of fissile isotopes in the form of a homogeneous solution in the moderator, with the fissile isotope selected from the group including ²³⁹ Pu and ²³⁵ U, total number of fissioning isotope less critical. 6. The multiplier according to claim 5, in which the concentration on the element of the fissile isotope is 1-100 g / l. 7. The multiplier according to claim 5, in which the moderator is water or heavy water, or graphite. 8. The multiplier according to claim 1, in which the neutron multiplication section and the neutron output section are each provided with a cooling system and a detonating gas conversion unit. 9. The multiplier according to claim 1, additionally containing a fuel insert placed vertically in the output compartment of the enhanced neutron flux and coaxially with it, and the height of the fuel insert is 0.5 of the average length ln of the neutron thermalization path in the fuel solution medium. 10. The multiplier according to claim 9, in which the height of the fuel insert in the output compartment of the enhanced neutron flux is chosen depending on the material of the fuel solution medium from the range of 2-4 cm when using light water (H ₂ O), 3-8 cm when using heavy water (D ₂ O), 8-11 cm when using graphite. 11. The multiplier according to claim 1, additionally containing a fuel insert placed at the output of the output compartment of the amplified neutron flux parallel to it and made in the form of a disk with a diameter equal to the mean path length ln of the neutron thermalization and 1-20 mm thick, is located at a distance of 0, 5 from the average length ln of the path of neutron thermalization in the medium of the fuel solution. 12. The multiplier according to claim 11, in which the diameter of the fuel insert at the output of the output compartment of the enhanced neutron flux is selected depending on the material of the fuel solution medium from the range of 5-7 cm when using light water (H ₂ O), 9-13 cm when using heavy water (D ₂ O), 17-21 cm using graphite.

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