RU2755261C1 - Nuclear power plant with ceramic fast neutron reactor - Google Patents

Abstract

FIELD: nuclear power plant.

SUBSTANCE: invention relates to a nuclear power plant. A nuclear power plant with a ceramic fast neutron reactor includes a nuclear reactor with fuel, fuel elements with casings, heat exchangers, pumps for circulating the coolant through heat exchangers with pipelines of the first, second and third circuits, and a generator. The reactor vessel, fuel element casings, internals, primary circuit pipelines, hot part of the secondary circuit pipelines, lithium-argon heat exchanger are made of inhibited composite ceramic material based on C-SiC or inhibited ceramic material based on SiC-SiC, argon-water heat exchanger is made of metal. Lithium (Li⁷) is used as a liquid heat carrier in the first loop, argon is used as a heat carrier in the second loop; uranium nitride or a mixture of uranium nitride with nitride and plutonium oxide in the form of tablets with a diameter of 10 to 40 mm and a height of 5 up to 100 mm.

EFFECT: comprehensive provision of safety enhancement in the event of a beyond design basis accident at a reactor, creation of a closed fuel cycle, an increase in the fuel reproduction ratio, a decrease in the specific metal consumption of a nuclear power plant, an increase in the nuclear power plant's operating life, an increase in the efficiency of a nuclear power plant, as well as the optimal choice of materials and structures of components and assemblies of a nuclear power plant.

1 cl, 1 dwg, 8 tbl

Description

The present invention relates to power engineering and can be used for the production of electricity and heat based on the use of nuclear fuel as an energy carrier. The invention relates to nuclear power plants (NPP).

The most efficient at present are fast reactors. A fast neutron power plant includes a nuclear reactor, a pump for circulating a liquid coolant through a heat exchanger, pipelines of the 1st, 2nd and 3rd circuits, a turbine, a generator and other equipment.

Nuclear power plants with pressurized water reactors on thermal neutrons are known from the state of the art (Margulova T.Kh. Electric stations. 5th ed. M .: MPEI, 1994, p. 21). NPPs of this type cannot generate steam with a high temperature and pressure close to the parameters achieved in traditional fossil fuel power generation. The parameters of such NPPs usually do not exceed 330 ° C and 7.0 MPa. The efficiency of water-moderated nuclear power plants reaches 35%. The disadvantages of NPPs with pressurized water reactors using thermal neutrons (VVER) are: adverse consequences in the event of a beyond design basis accident (taking into account the reaction of zirconium with water, since hydrogen and a huge amount of heat are released during the reaction); the impossibility of raising the temperature of the fuel pellet made of uranium oxide, due to the low coefficient of its thermal conductivity (the center of the pellet will overheat to unacceptable temperatures); the impossibility of implementing a closed fuel cycle (combustion in the reactor of only the isotope U ²³⁵ ); low efficiency of the NPP.

It is known that the cladding of fuel elements in VVER is made of an alloy of zirconium with niobium. At a temperature of 1200 ° C, which is reached in the process of residual heat release in the event of a beyond design basis accident, an exothermic reaction will occur between zirconium and water:

Zr + 2H ₂ O = ZrO ₂ + 2H ₂ ↑ + 6530 kJ / kg.

VVER-1000 was selected for the calculation. Taking into account the mass of zirconium 22,600 kg in the reactor, 993 kg of hydrogen will be released, 148 GJ of heat due to the substitution reaction and 140 GJ of heat due to the combustion of hydrogen in air. The released energy is equivalent to the explosion energy of 69 tons of TNT. In the event of a beyond design basis accident, the containment shell of the reactor shop will be destroyed and the area will be significantly contaminated with fission products. The scenario was partially realized at nuclear power plants in the cities of Fukushima (Japan), Chernobyl (USSR) and Three Mile Island (USA).

The closest to the declared technical solution is the BN-800 reactor (the fourth unit of the Beloyarsk NPP). Its characteristics: efficiency 39.4%, pressure in the 1st and 2nd circuits close to atmospheric, coolants of the first and second circuits - sodium, temperature at the inlet to the heat exchanger of the primary circuit 547 ° С, temperature at the exit from the heat exchanger of the second circuit 505 ° С, pressure in the third circuit is 140 atm, the coolant of the third circuit is water. Internals of the BN-800 reactor are made of stainless steel and zirconium. The reactor operates on fast neutrons, therefore, combustion in the reactor, including the isotope U ²³⁸ , is possible; it becomes possible to implement a closed fuel cycle. (Wikipedia, Rosatom website, Textbook for students "Fast neutron reactors" GB Usynin, EV Kusmantsev. NPP is indicated - unit 4 of the Beloyarsk NPP).

Disadvantages of this technical solution: the use of sodium as a heat carrier of the first and second circuits, which is a fire hazardous metal (the metal is very aggressive, when it interacts with water, hydrogen is released, and spontaneous combustion occurs); the impossibility of increasing the temperature of the primary circuit to high temperatures, since the boiling point of sodium is 883 ° C; the problem of maintaining the integrity of fuel elements in the presence of large fluins, which leads to the impossibility of large combustion of fuel with the existing design; low burn-up percentage of nuclear fuel containing several types of oxides of fissile materials Mixed-Oxide fuel (MOX) fuel; low reproduction rate; high specific indicator of metal consumption; limited service life of the NPP; low efficiency of the NPP.

The tasks to be solved by the claimed invention are to improve safety in the event of a beyond design basis accident at a reactor; in creating a closed fuel cycle; with increased fuel reproduction ratio; in reducing the specific metal consumption of nuclear power plants; in increasing the service life of nuclear power plants; in increasing the efficiency of nuclear power plants.

These tasks are solved by the proposed invention.

A nuclear power plant with a ceramic fast neutron reactor, including a nuclear reactor with fuel, a reactor vessel, fuel elements with jackets, heat exchangers, pumps for circulating a coolant through heat exchangers with pipelines of the first, second and third circuits, a generator, characterized in that the reactor vessel, covers of fuel elements, internals, primary circuit pipelines, hot part of secondary circuit pipelines, lithium-argon heat exchanger are made of inhibited composite ceramic material based on C-SiC or inhibited ceramic material based on SiC-SiC, argon-water heat exchanger is made of metal; ^{lithium (Li 7} ) is used as a liquid heat carrier in the first loop, argon is used as a heat carrier in the second loop; uranium nitride or a mixture of uranium nitride with nitride and plutonium oxide in the form of tablets with a diameter of 10 to 40 mm and a height of 5 to 100 mm.

The proposed ceramic fast neutron reactor (KRBN) has a number of advantages over atomic fast neutron reactors using uranium oxide and carbide as fuel, sodium and lead as the primary coolant, and stainless steel as the walls of the fuel elements. ...

In order to increase the world's energy security, nuclear power should use new generation technologies and a closed fuel cycle with fast breeder reactors. This will ensure the unlimited resource of nuclear fuel due to the transition from the use of the U ²³⁵ isotope to the use of natural uranium, thorium and plutonium. It is known that on the globe U ²³⁸ is 143 times more than U ²³⁵ . A huge amount of depleted uranium is waiting for its time to be involved in economic circulation.

Fast neutron reactors in the process generate not only energy, but also plutonium for nuclear weapons.

Fast reactors can generate electricity, heat for heating buildings and structures, and can also destroy long-lived high-level waste and produce materials for fuel reproduction.

The most optimal direction for the development of nuclear energy is the creation of a KRBN. With the transition from fast reactors with a sodium coolant (BN) to ceramic fast reactors (KRBN), it is possible to achieve improved technical and economic indicators.

The set tasks are solved due to the fact that the "ceramization" of the reactor will raise the temperature of the protective cover of the fuel element from 720 ° C to 1250 ° C, and the temperature of the coolant at the outlet of the reactor - from 540 ° C to 1200 ° C. The argon temperature at the exit from the heat exchanger will be about 1150 ° C, which will lead to an increase in the KRBN efficiency from 40% to 75%.

The design of the KRBN will make it possible to reduce the cost of construction, increase the unit capacity of the reactor and the main components of the power unit.

The elimination of zirconium, sodium, lead and water from the inside of the reactor space will increase the safety of the nuclear power plant. As a nuclear fuel instead of uranium oxide or carbide, fuel based on uranium nitrides or uranium and plutonium nitride is proposed. Instead of fire hazardous sodium and high-temperature-molten lead, Li ^{7 is} proposed as the primary coolant. By dividing natural lithium into Li ⁶ and Li ⁷ , the first isotope will be expediently used in thermonuclear power engineering and for defense purposes, and the second isotope - in nuclear power engineering.

The high nuclear density of uranium nitride (UN), when used, has an advantage over uranium carbide and uranium oxide fuels. The use of UN increases the efficiency of the primary fuel and the breeding ratio of the secondary fuel.

A high thermal conductivity UN will allow increasing the thermal loads in the fuel elements, as well as the size of the fuel pellets, which will contribute to a lower cost of manufacturing fuel elements and reduce the loss of neutrons in structural materials.

Uranium mononitride and mixed uranium and plutonium mononitrides, which have a number of favorable physical properties, are potentially important types of nuclear fuel and fertile materials. They have high dimensional stability during irradiation, and their use in nuclear reactors allows deep burnup to be achieved and, therefore, reduce the cost of the nuclear fuel cycle. The characteristic features of simple and mixed nitrides of uranium and plutonium are: high, in comparison with oxides, thermal conductivity; increased density and better ability to retain gaseous fission products (GPG).

The temperature of the UN fuel is lower than the temperature of the oxide fuel at the same volumetric energy release, which makes it possible to reduce the release of GPE during irradiation.

The operating temperature of nitride fuel is well below its permissible operating temperature limit, which leads to a potential increase in the level of safe operation due to the lower value of the negative Doppler effect.

Better retention of the HPA of the UN fuel reduces the amount of HPA in the fuel-cladding gap and reduces the gas pressure under the cladding of the fuel element.

The higher density of UN fuels as compared to oxide and carbide fuels, at lower enrichment, can lead to faster expanded breeding rates, shorter doubling times, and longer fuel campaign times.

The compatibility of UN fuel with a lithium coolant increases the safety of the KRBN operation.

There is a possibility of manufacturing not cylindrical fuel rods, but manufacturing fuel rods, for example, of a hexagonal shape, while it is possible to increase the fuel: coolant ratio, and this, in turn, will lead to the creation of a harder neutron spectrum in the reactor and, accordingly, their less “inappropriate” use. therefore, the reproduction rate will be maximum.

The physical and mechanical properties of uranium oxide, nitride and carbide are shown in Table 1.

Table 1

Fuel types Density, g / cm ³ Thermal conductivity, W / m * K Strength, MPa Melting point, ° С α
10 ^-6 C ^-1 Uranium oxide
UO ₂ 10.9 8.4 at 45 ° C
2.6 at 1400 ° C 8 - tensile 2750 ten Uranium nitride
UN 14.32 24 at 1200 ° C 15 - tensile 2850 ten Uranium carbide UC 13.63 32.7 at 45 ° C
23.6 at 2300 ° C 13 - tensile 2400 ten

The reaction of interaction of zirconium with water is excluded due to the absence of these substances in the core. The use of silicon carbide as a material for the wall of the fuel element cover will make it possible to get rid of the zirconium alloy in the interior of the reactor space.

Fuel-element claddings and internals in reactors (fuel rod and fuel rod ends, spacer grids of fuel assemblies (FA), a tube for the assembly of in-core detectors (SVRD), a central tube and a guide channel) are proposed to be made from a material based on inhibited materials of the C-SiC and SiC-SiC systems.

The technical result achieved is that uranium nitride UN is characterized by a much lower radiation swelling at temperatures of 1200-1700 ° C than uranium carbide (UC) and uranium oxide (UO ₂ ). The outgassing of fission products from UN fuel pellets is an order of magnitude less than from UC pellets. UN is not a pyrophoric material, unlike UC.

Uranium nitride fuel is technologically advanced during processing, since UN is completely soluble in nitric acid. The design of the fuel elements changes - the diameter of the fuel pellets and their height increase, which will lead to a smaller number of them, therefore, to a cheaper manufacture, as well as an increase in the fuel reproduction ratio.

Due to an increase in the temperature of the peripheral surface of the uranium nitride pellet in comparison with the temperature of the peripheral surface of the pellet made of uranium oxide, due to a sharp increase in the light component, the heat transfer to the fuel element cladding will significantly increase, despite an increase in the gap between the peripheral surface of the pellet and the inner wall of the fuel element cladding. This will allow, without loss of thermal efficiency, to sharply increase the volume for the accumulation of gases formed during the fission of nuclear fuel, and will not allow creating a high pressure of gases inside the fuel element. The possibility of swelling of the fuel without the possibility of contact with the protective cover will also remain. An increase in the diameter of fuel pellets and, consequently, a change in the "fuel-coolant" proportion when using Li ⁷ as a coolant, will preserve the "fast" spectrum of neutrons in the reactor. The optimum tablet height is in the range of 5-100 mm. If the height is less than 5 mm, then this will lead to an increase in the number of tablets, and tablets with a height of more than 100 mm are not feasible. The optimum tablet diameter is 10-40 mm. If the diameter is less than 10 mm, this will lead to an increase in the number of tablets, and in tablets with a diameter of more than 40 mm, the center of the tablet will overheat.

The proposed materials as a reactor vessel, fuel element covers, internals, primary piping, hot part of secondary piping, heat exchanger are made of materials based on inhibited C-SiC and SiC-SiC systems.

Inhibited materials are used to increase the service life. These materials have high chemical resistance to H ₂ O, Na, K, Pb, Li, etc .; high radiation; retain high strength up to a temperature of 1500 ° C; it is possible to make gas-tight cladding of fuel elements from them; SiC has a high melting point of 2730 ° C.

The neutron capture cross section for silicon carbide is orders of magnitude smaller than the neutron capture cross section for zirconium indicated in Table 2.

table 2

Element Neutron capture cross section,
barn Zirconium 5 Silicon 0.13 Carbon 0.003

The efficiency of the use of neutrons in KRBN sharply increases.

The creation of a closed fuel cycle will make it possible to use U ²³⁸ in spent nuclear fuel and in depleted uranium hexafluoride remaining after enrichment. Depleted uranium already produced will last for many hundreds of years.

The use of KRBN will make it possible to increase the depth of nuclear fuel burnup and bring the breeding ratio to 1.45.

KRBN will provide a new level of environmental safety due to a multiple reduction in the volume of spent nuclear fuel and radioactive waste. The waste will be “incinerated” in the KRBN. KRBN will increase the service life of nuclear power plants up to 100 years, since ceramic composite materials do not rust and practically do not degrade during irradiation.

The use of Li ⁷ as the primary coolant will allow raising the coolant temperature at the reactor outlet to 1200 ° C (the boiling point of lithium is 1330 ° C). Mercury, potassium and sodium have a boiling point significantly below 1200 ° C, respectively, these materials cannot be used in KRBN at high temperatures.

Lead-bismuth eutectic has a disadvantage due to the conversion of bismuth when interacting with neutrons into highly radioactive polonium. Lead requires high energy consumption for its pumping in the primary circuit. Lead oxides actively interact with many materials, including ceramic, which makes it difficult to use lead as a primary coolant.

Li ⁷ has a small neutron capture cross-section, which has a positive effect on the nuclear characteristics of the reactor. Li ⁶ remaining after the isotopic separation of lithium is expedient to be used for defense purposes and in thermonuclear power engineering of the future.

The low melting point of lithium (180 ° C) and its high boiling point (1330 ° C), low vapor pressure have a positive effect on the safety of the reactor.

The problem of heating a large amount of lead to melt it before starting the reactor is a complex technical problem (the melting point of lead is 327 ° C).

Due to the high thermal conductivity of lithium, the permissible heat fluxes practically do not limit the critical heat loads.

Lithium is a monoatomic metal, so there are no problems with radiation damage in the coolant. Li ⁷ in the reactor is not activated. Lithium is a more inert material and less fire hazardous than sodium and potassium. The high electrical conductivity of lithium allows the use of hermetically sealed AC and DC electric pumps. Lithium is a relatively cheap metal. The thermal conductivity of lithium is 3.5 times higher than the thermal conductivity of lead, its heat capacity is 27 times that of lead, besides, lithium is more widespread in the earth's crust than lead, which is an advantage when using lithium as a primary coolant. Lithium has an exceptionally high heat capacity by weight in comparison with other liquid metal coolants. The approximate heat capacity of lithium is:

t⁰C - 180 + 427 Cρ = 1.224 kcal / kg⁰C

t⁰C - 180 + 1227 Cρ = 1.045 kcal / kg⁰C

and, accordingly, lithium has a higher volumetric heat capacity.

The volumetric heat capacity of molten metals is presented in Table 3.

Table 3

Heat carrier C _ργ [kcal / m ³ ° C] t = 300 ° C t = 500 ° C Lithium 618 505 Sodium 274 250 Alloy Na-K 169 157 Pb-Bi alloy 366 354

With the same hydraulic resistance of the circuit, taking into account the low density of lithium, the lithium velocity is approximately equal to the Na or Na-K velocity and is approximately 4 times higher than that of the Pb-Bi eutectic melt.

Lithium has a large value of C _ργ · W in comparison with other heat transfer fluids, where (C _{ργ is the} heat capacity, W is the speed), given in Table 4.

Table 4

Heat carrier C _ργ · W [kcal / m * s ° C] t = 500 ° C Lithium 3530 Sodium 1750 Alloy Na-K 1100 Pb-Bi alloy 620

The value of C _ργ W is extremely important in the design of nuclear power transport installations due to the following circumstances:

a) in proportion to the value of C _ργ · W, it is possible to reduce the flow cross-sections of the entire primary coolant path and, therefore, to reduce the size and weight, as well as to simplify the layout of the equipment.

Lithium makes it possible to reduce the flow area of the primary circuit with the same reactor power and the same heating approximately 2.2 times compared to sodium, 3.6 times with Na-K alloy, 6.7 times with Pb-Bi eutectic.

A large volumetric heat capacity for lithium allows one to proportionally reduce the flow rate of the coolant and reduce the energy consumption for pumping the coolant in the primary circuit;

b) the large value of C _ργ W for Li is especially important for heat removal directly in the reactor, since the flow area for the coolant in the core can be _{directly proportional to C ργ} occupied by the moderator. The transition to a lithium coolant will, obviously, due to this, large savings in the loading of U ²³⁵ in comparison with sodium or lead-bismuth coolant.

A decrease in the flow area for the coolant in the end screens (the presence of a coolant flow in the end screens is characteristic of almost all reactor designs) gives an improvement in the end screens, which will have a favorable effect on the U ²³⁵ loading, as well as on the uniformity of the heat release field in the core;

c) while maintaining the same flow area in the reactor core for various liquid metal coolants in the case of using a lithium coolant, it is possible to reduce the temperature difference between the coolant inlet and outlet in proportion to C _ργ W (2.2 times compared to Na, 6, 7 times compared to Pb-Bi). At the same average outlet temperature of the coolant, this significantly increases the mean log temperature differences between the primary and secondary coolants. This makes it possible to significantly reduce the heat exchange equipment, i.e. significantly reduce the dimensions and weight of the installation;

d) at the same time, a decrease in the temperature difference of the coolant between the inlet and outlet leads to a decrease in the maximum temperature on the wall of the fuel element at the same average outlet temperature, since the excess of the maximum outlet temperature over the average is proportional to ∆t ° C. This circumstance is of great importance, since a good leveling of the energy release field, especially for small-sized homogeneous power devices, is very important;

e) a large value of C _ργ · W for lithium gives more space in the design of fuel elements than for other liquid metals, since it allows one to consider fuel elements of the Field type.

Lithium has a high thermal conductivity in a temperature range of interest for reactor design. The high thermal conductivity of lithium in combination with a high heat capacity by weight provides good heat transfer coefficients to lithium. Table 5 provides a comparison for the heat transfer coefficient in relation to lithium, sodium and Pb-Bi at speeds of Li and Na W = 7 m / s and a speed of Pb-Bi W = 2 m / s in kcal / m · h · ° C.

Table 5

Heat carriers t = 300 ° C t = 500 ° C d ₂ = 5 · 10 ^-3 m 10 · 10 ^-3 m 5 · 10 ^-3 m 10 · 10 ^-3 m Lithium 63100 48,300 61,700 45,600 Sodium 58200 39,200 49,200 33,400 Eutectic Pb Bi 15200 11,000 17,300 12100

Large heat transfer coefficients for Li provide small temperature differences between the surface of the fuel element and the coolant and lead to a decrease in the maximum temperature of the fuel element. Large heat transfer coefficients for lithium make it possible to reduce the difference between the temperature of the cladding of fuel elements and lithium.

Lithium has a fairly good electrical conductivity, and for lithium, as well as for sodium, electromagnetic pumps with a sufficiently high efficiency can be created. Considering the lower required costs for a given installation, an electromagnetic pump for lithium will be simpler.

Lithium compares favorably with sodium in that Li ⁷ , the use of which as a coolant must be guided by, does not give γ-activity when it is irradiated with neutrons. As a result, the lithium circuit does not require the powerful protection required in the case of sodium, which significantly increases the weight of the entire power plant.

Lithium, having a low specific gravity, even when using large volumes of it, does not give a noticeable increase in the weight of the installation. This fact is of no small importance in comparison with the use of Pb-Bi as a heat carrier.

Lithium is the most inactive element of all alkali metals. A "lithium" power plant will be more reliable and safe than using sodium or Na-K as a heat carrier.

Lithium, unlike Pb-Bi alloy, will not give α-active aerosols, which makes it possible to simplify the design and operation of the installation. Lithium does not interact with inhibited silicon carbide at temperatures up to 1300 ° C. Lithium has a very high latent heat of vaporization, as shown in Table 6.

Table 6

t, ° С Pressure, atm
R Latent heat of vaporization of 1 kg of lithium,
r kcal / kg 1336 1 4680 1540 ten 4100 1620 25 3800 1682 50 3500 1718 75 3300

A significant amount of energy is required to vaporize lithium, which also increases the safety of nuclear power plants.

The physical properties of liquid metal coolants sodium, lead and lithium are shown in Table 7.

Table 7

Options Sodium Lead Lithium
at 527 ° C Density at 450 ° С, kg / m3 845 10470 483 Temperature, ° С
melting
boiling 97.8
883 327.4
1737 180
1330 Thermal conductivity at 450 ° C, W / (m K) 68.9 15.7 54.4 Heat capacity at 450 ° C, kJ / (kg K) 1.272 0.155 4.181 Distribution in the earth's crust,% 2.85 1.6 · 10 ^-3 2.1 · 10 ^-3

The use of argon as a coolant in the secondary circuit will increase the safety of nuclear power plants, since the possibility of contact of an alkali metal with water is excluded, in contrast to a fast neutron reactor with a sodium coolant with an electric power of 880 MW (BN-800).

The use of a gas turbine in comparison with a steam turbine will reduce the size of the turbine; make a faster start of the turbine; increase the durability of the turbine due to less vibration, as well as fewer rubbing parts; reduce metal consumption; increase efficiency.

Ceramic composite material based on inhibited C-SiC and SiC-SiC retains high strength at temperatures up to 1500 ° C, radiation-resistant, inert to the melt of lithium, argon and other materials.

To increase the efficiency of the nuclear power plant, it is possible to heat buildings and structures with a third circuit, which uses water as a heat carrier.

The high-energy neutron flux in the KRBN is able to effectively “burn” the most dangerous long-lived radionuclides formed in the spent nuclear fuel. By using a closed fuel cycle with actinide burning and transmutation of long-lived fission products into short-lived ones, it is possible to radically solve the problem of nuclear power waste disposal and greatly reduce the volume of radioactive waste to be buried.

The transition to fast breeder reactors along with thermal reactors, as well as the transition to a closed fuel cycle, will make it possible to create a safe energy production technology that fully meets the requirements of sustainable development of human society.

The technical and economic indicators of the BN-1200 and KRBN-1000 reactors are shown in Table 8.

Table 8

Technical and economic indicators Reactor BN-1200 Reactor KRBN-1000
(calculated) Installed capacity utilization factor 0.9 0.95 Efficiency 45 75 Gradual increase in MOX fuel burnup from the achieved level twenty% 40% Reproduction rate 1,2 1.45 Decrease in specific indicators of metal consumption 1.7 times 3 times Reactor life 60 100

The development of the KRBN will lead to the solution of the set goals.

The figure shows a general diagram of a nuclear power plant, which consists of a reactor 1, fuel elements 2, an electric pump 3, a lithium-argon heat exchanger 4, a reactor vessel 5, internals 6, primary circuit pipelines 7, a hot part of a secondary circuit pipeline 8, a gas turbine 9, generator 10, cold part of the pipeline of the second circuit 11, argon-water heat exchanger 12, electric pump of the second circuit 13, pipeline of the third circuit 14, pump for water circulation in the third circuit 15.

The NPP operates with a ceramic fast neutron reactor (1) as follows. Uranium 238 and plutonium fission when fast neutrons are absorbed in tablets of uranium nitride or a mixture of uranium nitride with nitride and plutonium oxide, while energy is released (the temperature of the tablets is about 1500 ° C in the center and 1300 ° C at the periphery). Then, the walls of the fuel elements (2), made of a material based on an inhibited composite ceramic material based on C-SiC or an inhibited ceramic material based on SiC-SiC, are heated; their temperature is about 1250 ° C. Then the heat is transferred to the liquid Li ⁷ coolant of the primary circuit, which heats up to 1200 ° C, Li ⁷ and is pumped by an electric pump (3) to the lithium-argon heat exchanger (4) and back to the reactor; the reactor vessel (5), fuel element covers, internals (6), primary circuit pipelines (7), hot part of the secondary circuit pipeline (8), lithium-argon heat exchanger are made of materials based on inhibited composite ceramic material based on C-SiC or inhibited ceramic material based on SiC-SiC. In a lithium-argon heat exchanger, lithium transfers heat to argon, which heats up to a temperature of 1150 ° C and is fed through the pipelines of the hot part of the secondary circuit to the gas turbine (9), where it gives energy to the turbine, while the generator (10) generates electricity. Then argon is directed through the cold part of the secondary circuit pipeline (11) to an argon-water heat exchanger (12), where it heats water to 90 ° C; then, argon is pumped over by an electric pump (13) again enters the lithium-argon heat exchanger, where it heats up again, and the heated water is supplied through the third circuit pipelines (14) to heat buildings and structures; argon-water heat exchanger and pipelines of the third circuit are made of metal; then water, after transferring heat to buildings and structures, the water is returned to the argon-water heat exchanger by means of an electric pump (15) for reheating; if necessary, it is possible to add water to the third circuit to compensate for its losses.

Thus, the optimal combination of reference and new solutions and the possibility of expanded reproduction of fuel make it possible to classify the KRBN-1000 project as a fourth-generation nuclear technology.

Claims (1)

A nuclear power plant with a ceramic fast neutron reactor, including a nuclear reactor with fuel, a reactor vessel, fuel elements with jackets, heat exchangers, pumps for circulating a coolant through heat exchangers with pipelines of the first, second and third circuits, a generator, characterized in that the reactor vessel, casings of fuel elements, internals, primary piping, hot part of secondary piping, lithium-argon heat exchanger are made of inhibited composite ceramic material based on C-SiC or inhibited ceramic material based on SiC-SiC, argon-water heat exchanger is made of metal; ^{lithium (Li 7} ) is used as a liquid heat carrier in the first loop, argon is used as a heat carrier in the second loop; uranium nitride or a mixture of uranium nitride with nitride and plutonium oxide in the form of tablets with a diameter of 10 to 40 mm and a height of 5 to 100 mm.

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