RU2769102C1 - Passive cooling system of a nuclear reactor - Google Patents

Abstract

FIELD: nuclear power industry.

SUBSTANCE: invention relates to nuclear shell-type reactors with natural circulation of a coolant under pressure. The core cooling system contains a reactor vessel with inlet and outlet pipes located in a pool of water, circulation pipelines, a volume compensator and a heat exchanger of the first cooling circuit located above the water level in the pool. The inlet pipe is connected to the pipeline of the coolant cooled in the heat exchanger and is located in the lower part of the reactor vessel, and the outlet pipe is connected to the pipeline of the coolant heated in the reactor and is located in the upper part of the reactor vessel. The sections of pipelines connected to the inlet and outlet pipes are equipped with passive opening valves to reduce pressure in the first circuit, providing a shortened natural circulation circuit. The height difference between the core and the primary circuit heat exchanger is selected from the condition of providing the required pressure. The final heat sink can be the natural environment, and to prevent the release of radioactive substances into the environment, an intermediate heat transfer circuit made in the form of a gravitational heat pipe is used.

EFFECT: providing physical protection of the reactor plant, increasing the reliability of the cooling system by eliminating possible emergencies associated with de-energization of circulation pumps, as well as with the failure of pumps and shut-off valves in the cooling circuit.

4 cl, 5 tbl, 3 dwg

Description

State of the art

In recent years, there has been a significant increase in interest in the development of commercially viable reactor plants that use the natural circulation phenomenon (also known as the thermosiphon effect) to provide a flow of primary coolant to cool the core of a nuclear reactor. The present invention relates generally to nuclear reactor systems, and in particular to nuclear reactor systems using natural circulation of single-phase primary coolant, such as pressurized water pressure vessel reactors.

Known nuclear power plant AST-500 (Russia) with a reactor with a capacity of 500 MW (thr.), designed to generate low-temperature heat for district heating and hot water supply of cities. AST-500 is a pressurized water reactor with an integral layout of the primary circuit elements and natural circulation of the primary coolant. The features of the AST-500 reactor include natural circulation of the primary coolant at reduced operating parameters and features of the integral reactor, such as a built-in steam-gas pressure compensator, in-reactor heat exchangers for emergency heat removal, an external heat exchanger, and a containment shell. The main disadvantage of this boiling water reactor is the uneven fuel burnup along the height of the core due to the increasing steam content, which reduces the economic performance of the plant. The disadvantages include the impossibility of using it for other purposes (production of isotopes, irradiation of materials, scientific research, etc.), except for the generation of low-potential heat.

Known power vessel reactors with boiling water under pressure and natural circulation of the coolant, for example, the VK-50 reactor (Russia) with an installed electrical power of 50 MW. The steam produced in the reactor core goes directly to the turbine, producing electricity. The disadvantages of this installation include the manifestation of a self-oscillating regime in terms of power and the uneven burnup of fuel rods along the height of the core.

Canadian patent CA 1070860 A describes a nuclear facility comprising a pressurized water pressure vessel (PWR) nuclear reactor, referred to as a fundamentally safe reactor. In accordance with this patent, the reactor vessel containing the core is made of steel and insulated from the outside. The reactor vessel, in the upper and lower parts of which systems of hydraulic seals are built, is immersed in a pool of water, which has its own vessel. The reactor vessel has an outlet pipe in the upper part for water, which passes through the core, is heated and, with the help of an appropriate supply pipe, is removed from the pool to the heat exchanger. From the heat exchanger, water is fed back through an appropriate return pipe to an inlet pipe located under the core in the reactor pressure vessel. A circulation pump is installed on the return pipe of the primary circuit. The reactor core, two branch pipes, a supply pipe and a return pipe with a circulation pump installed on it, as well as a heat exchanger form the first circuit of the reactor.

In the aforementioned Canadian patent, fundamental safety is ensured by the fact that the water in the pool is under pressure and a connection device is provided that, in the event of an accident, ensures the free flow of water from the pool into the lower connection on one side, as well as a connection device that ensures the free flow of water from the upper branch pipe to the pool on the other side. A possible accident can be expressed in the fact that if the primary circuit pump fails, this will lead to an increase in the temperature inside the reactor.

The device for connecting the bottom connection to the water in the pool is an air seal or even an open pipe, into which, under normal operating conditions, no flow enters due to the appropriate selection of pressures, as explained below. The device for connecting the upper pipe to the water in the pool is a bell of gas or compressed steam installed in the upper part of a relatively high chamber, also filled with gas or steam. The height of said chamber must be such that the appropriate level of the liquid contained in the basin provides a pressure equal to the pressure drop of the water circulating in the reactor circuit. Thus, the lower nozzle of the reactor and the surrounding water in the pool are under the same pressure, i.e. there is no pressure difference between the two volumes: despite the fact that these two volumes communicate freely, the level of fluid flow between them is zero, since the fluids in them are under the same pressure.

In case of failure of the circulation pump, the pressure drop between the lower and upper pipes disappears, the pressure in the upper pipe increases and the water from the reactor is pushed into the chamber filled with gas, and from it into the pool. At the same time, water from the pool enters the lower branch pipe and from it enters the core area. The water from the reactor is thus replaced by water from the pool, which is colder, it has already been mentioned above that the reactor vessel is insulated. In addition, the water in the pool contains boron in such a way that, when it enters the reactor core, it gradually stops the reaction.

The volume of water in the pool is relatively large, which ensures a sufficiently large number of hours of operation of the reactor in the event of a failure of the primary circuit pump without overheating the core above a predetermined safe level.

From a purely technical point of view of operation, a fundamentally safe reactor of the type described above, which is the subject of Canadian patent CA 1070860 A, is not something exceptional. However, this known reactor has the disadvantage that, in the case of using a high temperature reactor, its design is too complex. Indeed, the pressure of the liquid in the pool must be higher than the pressure corresponding to the saturation temperature of the liquid leaving the core area, so either the amount of water in the pool is limited, then the shutdown of the reactor will be ensured, but cooling of the core is achieved only for a short time. , or the amount of water in the pool will be large, then a complex structure of reinforced concrete is required to ensure the retention of liquid under pressure.

Of the research reactors operating on natural circulation, a research-type nuclear reactor is known, described in patent RU2501103C1. The reactor core and reflector cooling system comprises a core and a reflector located in the reactor pool filled with coolant. The active zone and the reflector are placed in a housing made in the form of a box with two shells and a lower support grid with holes. The active zone is located in the inner shell of the body, and the reflector is located in the outer shell. The height of the inner shell is chosen from the condition of ensuring such a flow rate of the coolant due to natural circulation, which ensures the cooling of the core without exceeding the allowable temperatures of the fuel element claddings. The disadvantage of such a reactor is the impossibility of providing a high power of the reactor, due to the low boiling point at low pressure in the core, degraded heat transfer at natural circulation rates achievable in the pool, and an increased output of radiolytic gases from the reactor pool.

The closest analogue, coinciding with the claimed invention in the largest number of essential features, is the "Universal system of passive heat removal from the active zone of a research reactor" [1]. The principal feature of the proposed passive system for heat removal from a 10 MW reactor is the absence in the cooling circuit of not only active elements, such as circulation pumps and valves, but also passive elements with moving parts in the circulation path, such as a check valve. The cooling circuit includes only vessels, pipelines and a heat exchanger. The absence of elements with mechanical moving parts can significantly reduce the likelihood of equipment failures and increase the reliability of the cooling system while reducing its cost. The versatility of the proposed system makes it possible to use it for a wide range of developed typical reactor plants of various capacities intended for research in various areas of research and applied work related to nuclear technologies.

The disadvantage of this system is that in case of accidents with a rupture of the LOCA circulation pipeline, the coolant circulation loop through the reactor core is broken, which, at high levels of residual energy release, creates a threat of overheating and destruction of the core with the release of radioactive substances into the premises and the environment.

The inventive passive heat removal system makes it possible to eliminate these shortcomings and, with a significant simplification of the design, to provide the required safety indicators with an increase in the reactor power.

The essence of the invention

The invention is aimed at creating an efficient system for heat removal from a nuclear reactor, built on passive principles of operation and having an increased level of safety in case of possible emergency situations.

To eliminate the shortcomings noted above, solve the set technical problem and achieve the technical result according to the proposed invention in the well-known [1] safe nuclear installation, the design improvement is that:

• the reactor vessel is placed in a light water pool, preferably deep below ground level;

• the height difference between the core and the primary circuit heat exchanger is sufficient to create a large natural circulation driving head;

• minimal hydraulic resistance is ensured in all sections of the primary circuit, including the reactor pressure vessel and heat exchanger;

• The coolant inlet pipe is located in the lower part of the reactor pressure vessel, and the heated coolant outlet pipe is in the upper part;

• Passive valves are installed on the sections of the pipeline adjacent to the upper and lower branch pipes, which open when the pressure in the primary circuit decreases and ensure the creation of a shortened natural circulation circuit in the core through the reactor pool;

• the final heat sink is either atmospheric air circulating through the exhaust ventilation pipe, or circulating water in an open reservoir, or a thermal energy consumer, or a combination of the above systems;

• The system of heat removal from the reactor plant should preferably include an intermediate circuit made according to the principle of a gravitational heat pipe, with boiling of the working fluid in the primary circuit heat exchanger and steam condensation in the heat exchanger of the final absorber;

• the volume compensator is connected to the upper point of the primary circuit and is equipped with an overpressure protection valve;

• the primary circuit is connected to a monte-jus located at a level not higher than the reactor pool and providing coolant intake from the upper part of the primary circuit when switching to a shortened natural circulation circuit in the reactor pool to ensure refueling of fuel assemblies with the reactor vessel cover removed;

• The reactor vessel at the level of the core can be made of a material (eg zircaloy) that weakly absorbs neutrons, and the neutron reflector can be located outside the reactor vessel in a light water basin;

• heavy water can be used as a neutron reflector, for example, in a zircalic tank with channels for irradiation;

• A natural circulation scheme can be used to cool the heavy water in the reflector, and the heat is removed by a heat exchanger to the final absorber, for example, to atmospheric air.

достигается через обеспечение разности высот h между теплообменником и активной зоной, определяемой из соотношенияSolving the problem of achieving the required reactor power

is achieved by providing a height difference h between the heat exchanger and the active zone, determined from the relation

where

- максимальная мощность реакторной установки, Вт;

- maximum power of the reactor plant, W;

- требуемый расход теплоносителя в первом контуре при мощности

, кг/с;

- required coolant flow rate in the primary circuit at power

, kg/s;

- потеря напора на реакторе, включая активную зону, при расходе теплоносителя в первом контуре

, Па;

- pressure loss in the reactor, including the core, with coolant flow in the primary circuit

, Pa;

- потеря напора на трубопроводе с нагретым теплоносителем при расходе теплоносителя в первом контуре

, Па;

- pressure loss in the pipeline with heated coolant at coolant flow in the primary circuit

, Pa;

- потеря напора на трубопроводе с охлажденным теплоносителем при расходе теплоносителя в первом контуре

, Па;

- pressure loss in the pipeline with cooled coolant at coolant flow in the primary circuit

, Pa;

- потеря напора на теплообменнике при расходе теплоносителя в первом контуре

, Па;

- loss of pressure on the heat exchanger at the flow rate of the coolant in the primary circuit

, Pa;

- разность между средними плотностями теплоносителя на подъемном и опускном участках первого контура;

- the difference between the average coolant densities in the upstream and downstream sections of the primary circuit;

g - free fall acceleration, m/s ² .

Achieving these goals is achieved by the fact that the location of the reactor deep underground provides advantages over the usual architecture of research reactor facilities:

- most scenarios of external impact on the reactor plant are excluded, so many emergency situations resulting from explosions, tornadoes, snow loads, aircraft crashes, etc. can be ignored, there is no need for expensive containment, the physical protection of the reactor plant is simplified;

- the absence of expensive and complex systems for heat removal from the core can drastically reduce the cost of the reactor plant;

- a large depth of immersion of the core under the ground allows you to create a simple and highly efficient system of natural circulation. At the same time, there are no visible restrictions on increasing the height of natural circulation and the corresponding increase in the driving pressure through the core;

- the reactor can operate equally safely and efficiently in a very wide range of capacities based on the tasks of irradiation;

- there is practically no problem of expensive dismantling of the reactor plant.

During operation of the reactor, automatic adjustment of the flow rate in the cooling circuits is ensured when the power level changes, and the absence of the possibility for personnel of erroneous or malicious actions to reduce the intensity of coolant circulation eliminates emergency situations with deterioration of heat removal from the core. In addition, automatic adjustment of the flow rate in the cooling circuits when the power level changes ensures the utmost simplicity of the reactor plant control and reduces the requirements for personnel qualification. Therefore, such a reactor can operate in countries where there are no personnel with extensive experience in managing reactor facilities, and the reactor facility can also be used for training purposes.

Relatively low velocities of the upward flow of the coolant in the core at high pressure provide a large margin before the onset of boiling on the surface of the fuel elements and do not create problems for the upper location of the controls and protection systems on the reactor head, which greatly simplifies their design. The absence of fittings and pumps in the cooling circuits ensures smooth changes in heat removal parameters and completely eliminates water hammer.

The inventive reactor plant heat removal system with shortened intra-basin circulation valves provides a quick and easy transition to core cooling with pool water, which makes it possible to carry out transport and reloading operations without problems with the reactor vessel cover open.

Description of drawings

The following description relates to the accompanying drawings which show, by way of non-limiting example, an embodiment of the invention and in which, in FIG. Figure 1 shows a three-loop diagram of a research reactor facility, the final heat sink of which is atmospheric air circulating in the exhaust ventilation pipe. On FIG. Figure 2 shows the scheme for cooling the core with pool water by a shortened natural circulation circuit through passive pressure reduction opening valves. On FIG. 3, as an example of a specific design, shows the view of the reactor core and the fuel assemblies used.

Shown in FIG. 1, the three-loop heat removal system of a research reactor with natural circulation includes a pool with demineralized water 1, in which a pressurized reactor vessel 2 is placed with a removable cover 3, on which drives 4 of the operating elements of the control and protection system (CPS) 5 are placed, which move in the reactor core 6. The coolant heated in the reactor is removed to the primary circuit heat exchanger 7 from the upper part of the vessel through pipeline 8, and the cooled coolant is returned to the lower part of the reactor vessel through pipeline 9. Near the inlet and outlet nozzles of the reactor under the water level in the pool passive valves 10 and 11 are installed, which open when the pressure in the primary circuit decreases and ensure the creation of a shortened natural circulation circuit of the core with pool water. During operation, these valves are kept closed by high pressure in the primary circuit. Before filling the primary circuit and raising the pressure, the locking element of these valves is mechanically cocked, and after reaching the required pressure, it is released, and it is kept from opening only due to the pressure drop in the primary circuit and in the reactor pool. The filling of the primary circuit is carried out by squeezing water out of the monte-jus 12 and blowing gases through the volume compensator 13 connected to the upper point of the primary circuit. The height difference h between the core and the primary circuit heat exchanger provides the driving head of natural circulation. To ensure the convenience of working with irradiation channels 14 located in the neutron reflector 15, the latter is located outside the reactor vessel in pool 1.

The second cooling circuit is made in the form of a gravitational heat pipe, in which the evaporation of the working fluid (for example, water) occurs in the shell-and-tube steam generator 7, which is also the heat exchanger of the primary circuit. The generated steam through the steam line 16 enters the air heat exchanger 17, and the resulting condensate flows down through the pipeline 18 back to the steam generator 7.

The air heat exchanger 17 is part of the third cooling circuit and transfers heat to the final heat sink - atmospheric air circulating due to natural convection in the exhaust ventilation pipe 19. The atmospheric air entering through the openings 20 passes through the air heat exchanger and heats up, condensing the steam of the second circuit. The heated air, getting into the exhaust ventilation pipe 19 due to the difference in air densities in the pipe and in the surrounding space, provides a driving head of natural circulation. Thus, a three-loop heat removal system from a nuclear reactor can operate on a passive principle of operation.

The heat exchanger of the third circuit can also give off its heat to an open reservoir, which in this case will be the final heat sink. This is a more economical option, however, there may be problems with cleaning salt deposits on the heating surfaces of the heat exchange tubes.

The final heat sink can be, for example, a heating plant, however, in this case, the circulation of the heat carrier in this circuit is provided forcibly (by pumps).

On FIG. 2 shows a diagram of carrying out transport and technological operations for reloading the reactor with fuel assemblies. Before refueling, the coolant level in the primary circuit is reduced and a transition is made to heat removal from the core through a shortened natural circulation circuit with pool water with valves 10 and 11 open. To do this, the gas pressure in the volume compensator 13 is relieved and after opening the stop valves, monjus 12 is filled. In this case, the pressure on the shut-off elements of valves 10 and 11 decreases and they fall down under their own weight, opening the channel for the passage of the coolant. Residual heat release in the core is transferred to the pool water and is removed from the surface through evaporation to the special ventilation system. This allows the reactor to be refueled with the vessel lid open. In order to less frequently perform the operation of decompression of the reactor vessel, not only the core is reloaded, but also the cells for fuel assemblies (FA) 21 in the upper part of the vessel. This allows fuel assemblies to be refueled in the core without decompression of the primary circuit, under pressure.

As a non-limiting example of a specific design, a three-circuit cooling system for a pressurized research reactor with a thermal power of 25 MW, with a pressure in the primary circuit of 6 MPa, is considered. Figure 3 shows the reactor core with a diameter of 490 mm, consisting of 30 fuel assemblies of the VVR-KN type (24 fuel assemblies with 8 tubular fuel rods 22 and 6 fuel assemblies with 6 tubular fuel rods 23). In the central channels of 5 fuel rod fuel assemblies 24, the working bodies of the control and protection system are located. The space between the housing and the fuel assembly, as well as the space between the central moderating cavity 26 and the fuel assembly, is filled with a material 25 that weakly absorbs neutrons (for example, zircaloy, aluminum, beryllium, etc.).

The main thermal and neutron-physical parameters of the considered research reactor facility with a capacity of 25 MW with the calculated characteristics substantiated below are given in Table 1.

Table 1. Main parameters of the reactor with natural circulation of the coolant

Reactor characteristics Meaning Reactor type Pressurized water reactor with intermediate neutron spectrum and central trap Power, MW 25 Maximum heat flux density, cm ^-2 s ^-1 1.17×10 ¹⁵ Fuel Uranium dioxide, 20% U-235 enrichment Core geometry Cylindrical shape with a neutron trap in the center Number of cells for fuel assemblies, pcs thirty TVS type, pcs VVR-KN of them 5-pipe, pcs 6 8-pipe, pcs 24 Number of cooling circuits 3 Coolant of the 1st circuit light water Core diameter, mm ∅480 Active zone height, mm 600 Coolant consumption of the I circuit, [t h ^-1 ] 463.3 Temperature at the inlet to the fuel assembly, [°C] 118 Temperature at the outlet to the fuel assembly, [°C] 164 Coolant heating in the core, [°C] 46 Maximum temperature of fuel rods, [°C] 242 Pressure loss in the active zone, Pa 14100 Pressure at the core outlet, [Pa] 5.5× ¹⁰⁶ Hydraulic diameter of circulation pipes [mm] 400 Height of the I circuit with natural circulation [m] 80

Table 2 presents the characteristics of fuel assemblies taken for the calculation analysis.

Table 2. Characteristics of fuel assemblies BBP-KN [2]

Parameter Meaning Enrichment 235U, % 19.7 Density of uranium, g×cm ^-3 2.8 Mass of 235U in fuel assemblies, g 8-pipe 250 5-pipe 199 Number of fuel rods 8-pipe eight 5-pipe 5 Fuel rod thickness, mm 1.6 Core thickness, mm 0.7 Shell thickness, mm 0.45 Heat transfer area
surface, m ² 1.34

The main design parameters of the heat pipe steam generator in the presence of air in the intermediate circuit are given in Table 3.

Table 3 Main parameters of the heat pipe steam generator (presence of non-condensable gases, conservative calculation)

Parameter Meaning Material Stainless steel Thermal conductivity of the material, [W m ^-1 K ^-1 ] sixteen Transmitted power, [MW] 25 Heating medium consumption, [t h ^-1 ] 463.3 Temperature at the inlet to the fuel assembly, [°C] 164 Temperature at the outlet to the fuel assembly, [°C] 118 Pressure in the heat pipe, [MPa] 0.17 Steam temperature in the heat pipe, [°C] 115 Steam consumption in the heat pipe, [t h ^-1 ] 69.12 Mean logarithmic temperature difference, [°C] 24 External diameter of heat exchange tubes, [m] 0.025 Wall thickness of heat exchange tubes, [m] 0.002 Estimated heat transfer coefficient, [W m ^-2 K ^-1 ] 1483 Estimated heat exchange surface, [m ² ] 700 Accepted heat exchange surface, [m ² ] 765 Heat exchange tube length, [m] nine Number of heat exchange tubes, [pcs] 1063 Coolant velocity in tubes, [m s ^-1 ] 0.37 Pressure drop along the primary circuit, [Pa] 600

To evaluate the parameters of the air heat exchanger, the modeling of air heating during its passage through the gaps between elliptical stainless steel tubes with a wall thickness of 1 mm, the parameters of which are presented in Table 4, was carried out.

Table 4 Main parameters of the air heat exchanger

Parameter Meaning Transmitted power, [MW] 25 Material Stainless steel Thermal conductivity of the tube material, [W m ^-1 K ^-1 ] sixteen Heat exchange surface, [ ^m2 ]: °° - air side 1086.4 °° - steam side 1030 Cross-sectional size of elliptical heat exchange tubes, [mm] 10.4×18.4×1.5 Tube length, [m] nine Number of tubes in a row (right and left) 1120 Number of layers in the heat exchanger, [pcs] ten Number of tubes, [pcs] 22400 Hydraulic diameter by air, [m] 0.038 Hydraulic diameter for steam, [m] 0.014 Steam / condensate consumption (at 25 MW), [t h ^-1 ] 69.1 Air consumption (at 25 MW), [t h ^-1 ] 739 Air heating, [°С] 33.2 Air pressure drop, [Pa] fifty

Table 5 shows the calculated parameters of the exhaust pipe with a height of 145 m, with a diameter of the passage section in the lower part of 12.6 m, and in the upper part of 10.4 m.

Table 5 Main design parameters of the exhaust pipe with a height of 145 m

Power output, MW ten 25 fifty 100 Inlet air temperature, °C 20 20 20 20 Outlet air temperature, °C 37.3 53.2 73.5 110 Air consumption, kg/s 571 739 932 1108 Pressure drop across the heat exchanger, Pa 33 fifty 69 100

List of sources

1. Vitaly Uzikov, Irina Uzikova, “Universal system of passive heat removal from the core of a research reactor”, Nuclear Technology and Radiation Protection., Vol. XXXV, no. 2 June 2019.

2. Arinkin F.M., Shaimerdenov A.A. et al. / Conversion of the active zone of the WWR-K research reactor, Atomic Energy, 2017, vol. 123, No. 1 - p. 15-20.

Claims (14)

1. Passive cooling system of a nuclear reactor in the form of a natural circulation circuit, including a vessel reactor located below ground level, a primary circuit heat exchanger located above the reactor, inlet and outlet circulation pipelines, and a volume compensator, characterized in that the vessel with the reactor core , as well as the inlet and outlet pipes with adjacent sections of the circulation pipelines, are located in the pool with water, the inlet pipe connected to the cooled coolant pipeline is located in the lower part of the housing, and the outlet pipe connected to the heated coolant pipeline is located in the upper part of the housing, sections of the circulation pipeline, adjacent to the nozzles, are equipped with passive opening valves to reduce the pressure in the primary circuit, and the height difference h between the heat exchanger and the core located under it in the reactor vessel is determined from the relation

where

- maximum power of the reactor plant, W;

- required coolant flow rate in the primary circuit at power

, kg/s;

- pressure loss in the reactor, including the core, with coolant flow in the primary circuit

, Pa;

- pressure loss in the pipeline with heated coolant at coolant flow in the primary circuit

, Pa;

- pressure loss in the pipeline with cooled coolant at coolant flow in the primary circuit

, Pa;

- loss of pressure on the heat exchanger at the flow rate of the coolant in the primary circuit

, Pa;

- the difference between the average coolant densities in the upstream and downstream sections of the primary circuit; kg / m ³ ; g - free fall acceleration, m/s ² . 2. Passive cooling system of a nuclear reactor in the form of a natural circulation circuit according to claim 1, characterized in that the primary circuit is connected to a monte-jus, which provides partial emptying of the primary circuit during core transport and reloading operations. 3. Passive cooling system of a nuclear reactor, made according to a three-circuit cooling scheme, in which the first circuit with natural circulation includes a pressure vessel reactor located below ground level, a heat exchanger that transfers heat from the first to the second circuit and is located above the reactor, inlet and outlet circulation pipelines and a volume compensator, characterized in that the final heat sink of the third circuit is the natural environment in the form of atmospheric air and/or open water, and the second circuit is made in the form of a gravitational heat pipe that transfers heat from the first circuit to the third one. 4. Passive cooling system of a nuclear reactor, made according to a three-circuit cooling scheme according to claim 3, characterized in that the natural circulation of the final heat sink in the form of atmospheric air through the air heat exchanger of the third circuit is provided by an exhaust pipe.

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