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WO2024131546A1 - 一种核反应堆 - Google Patents

一种核反应堆 Download PDF

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Publication number
WO2024131546A1
WO2024131546A1 PCT/CN2023/137013 CN2023137013W WO2024131546A1 WO 2024131546 A1 WO2024131546 A1 WO 2024131546A1 CN 2023137013 W CN2023137013 W CN 2023137013W WO 2024131546 A1 WO2024131546 A1 WO 2024131546A1
Authority
WO
WIPO (PCT)
Prior art keywords
pipe
cavity
heat exchange
reactor vessel
exchange device
Prior art date
Application number
PCT/CN2023/137013
Other languages
English (en)
French (fr)
Inventor
梁活
林继铭
段承杰
崔大伟
宋磊
沈永刚
王迪
徐伟峰
王晓婷
徐昌恒
潘晖
廖子昱
袁昭君
Original Assignee
中广核研究院有限公司
中国广核集团有限公司
中国广核电力股份有限公司
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by 中广核研究院有限公司, 中国广核集团有限公司, 中国广核电力股份有限公司 filed Critical 中广核研究院有限公司
Publication of WO2024131546A1 publication Critical patent/WO2024131546A1/zh

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Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactor types
    • G21C1/32—Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactor types
    • G21C1/02—Fast fission reactors, i.e. reactors not using a moderator ; Metal cooled reactors; Fast breeders
    • G21C1/03—Fast fission reactors, i.e. reactors not using a moderator ; Metal cooled reactors; Fast breeders cooled by a coolant not essentially pressurised, e.g. pool-type reactors
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
    • G21C15/02—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices
    • G21C15/14—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices from headers; from joints in ducts
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
    • G21C15/18—Emergency cooling arrangements; Removing shut-down heat
    • G21C15/182—Emergency cooling arrangements; Removing shut-down heat comprising powered means, e.g. pumps
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
    • G21C15/24—Promoting flow of the coolant
    • G21C15/243—Promoting flow of the coolant for liquids
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors

Definitions

  • the present application relates to the technical field of nuclear power equipment, and in particular to a nuclear reactor.
  • the international community has carried out research on small mobile nuclear reactors with inherent safety, simple systems, and the ability to be fully prefabricated and quickly assembled in factories.
  • the containment of the small mobile nuclear reactor is set to a cylindrical shape. The resulting problem is that the height difference between the reactor core and the heat exchanger is small, and the circulation capacity is weak.
  • a nuclear reactor comprising:
  • a reactor vessel wherein a liquid coolant is arranged in the reactor vessel, wherein the liquid coolant is located in the reactor vessel The portion above the coolant forms an air cavity;
  • the heat exchange device being arranged in the reactor vessel and below the liquid level of the liquid coolant, the heat exchange device comprising an inner wall and an outer wall, the inner wall surrounding a first cavity for accommodating a core, and the outer wall and the inner wall jointly defining a closed second cavity;
  • a driving device is arranged in the reactor vessel and is located higher than the liquid level of the liquid coolant, and comprises a suction end and a re-injection end, the suction end is communicated with the air cavity, and the re-injection end extends into the first cavity and is arranged close to the core;
  • a water supply device is connected to the second cavity and is used to provide circulating cold source water required by the heat exchange device.
  • a plurality of heat dissipation fins are disposed on opposite side walls of the second cavity, the plurality of heat dissipation fins are circumferentially spaced along the side walls of the second cavity, and the heat dissipation fins extend vertically along the side walls of the second cavity.
  • a plurality of support members are arranged inside the reactor vessel, including a first support member arranged on the inner side wall of the reactor vessel and a second support member arranged on the bottom of the reactor vessel, and the heat exchange device is suspended in the reactor vessel in cooperation with the support members.
  • the liquid coolant is a molten metal coolant
  • the air cavity is filled with an inert gas
  • the driving device includes a gas compression pump, one end of which is provided with a suction pipe connected to the air cavity, and the other end is provided with a reinjection pipe, an annular injection pipe is provided near the core, and a plurality of upward nozzles are provided along the circumference of the annular injection pipe, one end of the reinjection pipe is connected to the gas compression pump, and the other end is connected to the annular injection pipe.
  • the reactor vessel includes a vessel body and a vessel cover, the upper end surface of the vessel body is provided with an opening, the opening is provided with a matching vessel cover, a steam generator is relatively provided on the inner wall of the vessel body, and the heat exchange device is provided on the steam generator so that the outer wall surface is close to the steam generator.
  • the heat exchange device further includes a coolant pipe, one end of the coolant pipe is connected to the water supply device, and the other end is connected to the second cavity, the coolant pipe includes an outlet pipe and an inlet pipe, and the outlet pipe is arranged above the inlet pipe.
  • the core includes a fuel rod bundle and a control rod assembly.
  • the nuclear power of the fuel rods undergoing nuclear reaction is controlled.
  • the outer side of the core is wrapped with a neutron reflection layer.
  • the water supply device is configured as a pool-type water supply device, including a water supply device for accommodating the reactor
  • the reactor container is provided with a pipeline through hole, and the outlet pipeline and the inlet pipeline are both connected with the pit through the pipeline through hole.
  • the water supply device is configured as a tubular water supply device, including an inlet mother pipe, an outlet mother pipe and a cold source water tank.
  • a plurality of the outlet pipes and a plurality of the inlet pipes are circumferentially arranged along the outer wall of the heat exchange device.
  • a plurality of the inlet pipes merge to form the inlet mother pipe, and a plurality of the outlet pipes merge to form the outlet mother pipe.
  • the inlet mother pipe and the outlet mother pipe are respectively used to connect the cold source water tank with the second cavity.
  • a second isolation valve is provided on the outlet mother pipe, and a third isolation valve is provided on the inlet mother pipe.
  • the height of the cold source water tank is higher than the reactor vessel, and an exhaust window communicating with the outside is provided on the upper end surface of the cold source water tank.
  • FIG1 is a front cross-sectional view of a nuclear reactor (excluding a water supply device);
  • FIG2 is a schematic diagram of the internal cross-section of a reactor vessel of a nuclear reactor
  • FIG3 is a front cross-sectional view of a heat exchange device of a nuclear reactor
  • FIG4 is a top cross-sectional view of a heat exchange device of a nuclear reactor
  • FIG5 is a schematic diagram of the structure of a nuclear reactor using a pool-type water supply device
  • FIG. 6 is a schematic diagram showing the structure of a tubular water supply device used in a nuclear reactor.
  • Cold source water tank 430. Drain pipe; 440. First isolation valve; 450. Exhaust window; 460. Inlet main pipe; 470. Outlet main pipe; 480. Second isolation valve; 490. The third isolation valve; 500, core; 510, neutron reflector.
  • the term "and/or" is only a description of the association relationship of associated objects, indicating that three relationships may exist.
  • a and/or B can represent: A exists alone, A and B exist at the same time, and B exists alone.
  • the character "/" in this article generally indicates that the associated objects before and after are in an "or" relationship.
  • multiple refers to more than two (including two).
  • multiple groups refers to more than two groups (including two groups), and “multiple pieces” refers to more than two pieces (including two pieces).
  • loop heat exchangers In the related art, nuclear reactors often use loop heat exchangers to form the power of natural circulation. The size of the power depends on the relative height difference between the heat source and the cold source. If the height difference is too small, natural circulation cannot be formed.
  • the use of loop heat exchangers requires the installation of heat exchangers as heat transfer interfaces. For small integrated reactors, there is no extra space to install heat exchangers.
  • loop heat exchangers use thin tubes with small diameters and thin walls, which are easy to break under high temperature and high pressure conditions and have low reliability.
  • the present application provides a nuclear reactor, including: a reactor vessel 100, a heat exchange device 200, a drive device 300 and a water supply device 400, wherein the reactor vessel 100 is provided with liquid coolant, and a portion of the reactor vessel 100 located above the liquid coolant forms an air cavity 150, and the heat exchange device 200 is provided in the reactor vessel 100 and is located below the liquid level of the liquid coolant, and the heat exchange device 200 includes an inner wall surface 210 and an outer wall surface 220, and the inner wall surface 210 is provided with a liquid coolant, and the outer wall surface 220 is provided with a liquid coolant, and the inner wall surface 210 is provided with a liquid coolant, and the inner ...20 is provided with a liquid coolant, and the inner wall surface 220 is provided with a liquid coolant, and the inner wall surface 210 is provided with a liquid coolant, and the inner wall surface 220 is provided with a liquid coolant, and the inner wall surface 220 is provided with a liquid coolant, and the inner wall surface 220 is provided
  • the nuclear reactor includes a reactor vessel 100, a heat exchange device 200, a drive device 300 and a water supply device 400, wherein liquid coolant is arranged in the reactor vessel 100, and the part located above the liquid coolant forms an air cavity 150, and the heat exchange device 200 is arranged below the liquid level of the liquid coolant in the reactor vessel 100.
  • the first cavity 240 surrounded by the inner wall surface 210 of the heat exchange device 200 is used to accommodate the core 500, and the outer wall surface 220 and the inner wall surface 210 jointly define a closed second cavity 250 connected to the water supply device 400.
  • the reactor core 500 When the nuclear power plant is operating normally, the reactor core 500 has a relatively high power.
  • the suction end of the driving device 300 extracts gas from the air cavity 150 and then injects it into the liquid coolant at the top of the core 500 through the reinjection end.
  • the injected gas makes the average density of the liquid coolant above the top of the core 500 less than the average density of the coolant near the steam generator 140 at the same height, thereby generating a gravity difference, driving the liquid coolant to circulate inside the reactor vessel 100, so as to form a circulation route from the outlet above the first cavity core 500-upper chamber 160-downward annular cavity 171-lower chamber 170-core 500 inlet-core 500-outlet above the core 500.
  • the liquid coolant flows from top to bottom in the reactor vessel 100 to transfer the heat generated by the core 500 to the steam generator 140 arranged in the downward annular cavity 171.
  • the driving device 300 stops operating.
  • the heat removal system configured for the core 500 cannot work properly, and the circulating cold source water in the water supply device 400 circulates in the second cavity 250 to continuously exchange heat with the core 500 to remove the excess heat of the core 500.
  • the reactor does not need to be equipped with a large-volume main pump, which reduces the complexity and cost of the entire system and avoids the problem of low system reliability caused by failure of the heat removal system.
  • the outer wall 220 is arranged outside the inner wall 210, and the heat exchange device 200 also includes a connecting wall 230 for defining the second cavity 250.
  • the connecting wall 230 is arranged between the outer wall 220 and the two end surfaces of the inner wall, forming an annular cavity sealed on all sides.
  • the inner and outer wall 220 of the heat exchange device 200 are in contact with the coolant of the core 500 of the first cavity 240, and the heat of the coolant can be transferred to the inner wall of the second cavity 250 and the heat dissipation fins 260 welded to the inner wall of the second cavity 250 by heat conduction.
  • the heat exchange device 200 in this embodiment is made of stainless steel material, and is used to bear the function of removing residual heat when the nuclear reactor is shut down.
  • the core 500 and the outer side are separated into two different circulation channels, so that the nuclear reactor does not need to be equipped with a core 500 hanging basket, saving equipment cost and space.
  • a plurality of heat dissipation fins 260 are arranged on the opposite side walls of the second cavity 250, and the plurality of heat dissipation fins 260 are arranged circumferentially at intervals along the side walls of the second cavity 250, and the heat dissipation fins 260 are extended in the vertical direction along the side walls of the second cavity 250.
  • the heat dissipation fins 260 are arranged as a thin sheet structure, which is used to increase the heat exchange area in contact with the cold source water and improve the heat exchange efficiency.
  • a plurality of support members are arranged inside the reactor vessel 100, including a first support member arranged on the inner wall of the reactor vessel 100 and a second support member arranged on the bottom of the reactor vessel 100.
  • the heat exchange device 200 is suspended in the reactor vessel 100 in cooperation with the support members to achieve a better heat exchange effect.
  • the heat exchange device 200 is suspended inside the reactor vessel 100 by a plurality of supports, and an upper chamber 160 is formed above the core 500 in the reactor vessel 100 and below the liquid coolant level, and a lower chamber 170 is formed below the core 500 in the reactor vessel 100, and the lower chamber 170 also includes a descending annular cavity 171 between the outer wall 220 of the heat exchange device 200 and the inner wall of the reactor vessel 100.
  • the liquid coolant is a molten metal coolant
  • the air cavity 150 is filled with an inert gas.
  • the molten metal coolant of the nuclear reactor can be a variety of different types of liquids, including liquid sodium, lead, lead-bismuth alloy, etc. Since the liquid metal coolant has a large density and good thermal conductivity, the use of molten metal coolant in the nuclear reactor can greatly increase the safety performance of the reactor.
  • the air cavity 150 at the liquid level of the molten metal coolant is filled with an inert gas to control the pressure of the molten metal coolant and keep it within a certain safety range.
  • the primary circuit uses molten metal as a coolant, which does not require too much high head difference between cold and hot sources, and can save layout space.
  • the driving device 300 includes a gas compression pump 310, one end of the gas compression pump 310 is provided with a suction pipe 320 connected to the air cavity 150, and the other end is provided with a re-injection pipe 330.
  • An annular jet pipe 340 is provided near the core 500 , and a plurality of upward nozzles 341 are provided circumferentially along the annular jet pipe 340 .
  • One end of the reinjection pipe 330 is connected to the gas compression pump 310 , and the other end is connected to the annular jet pipe 340 .
  • the gas compression pump 310 is set at a height higher than the liquid level of the liquid coolant inside the reactor vessel 100, that is, the positions where the suction pipe 320 and the re-injection pipe 330 are inserted through the wall of the reactor vessel 100 are both higher than the liquid level of the internal coolant.
  • the re-injection pipe 330 includes a first pipe body 331 inserted through the reactor vessel 100 and a second pipe body 332 that turns downward and extends downward along the inner wall 210 of the heat exchange device 200.
  • the second pipe body 332 is connected to the annular jet pipe 340, and multiple nozzles 341 distributed at intervals in the annular jet pipe 340 can discharge gas more evenly.
  • a gas compression pump 310 with a simpler configuration is used instead of the main pump. By adjusting the flow rate of the gas compression pump 310 to change the gravity difference, the coolant flow rate of the core 500 is changed, so that the flow path of the entire nuclear reactor is simple, the resistance generated is reduced, and the natural circulation capacity is improved.
  • the reactor vessel 100 includes a container body 110 and a container cover 120, the upper end surface of the container body 110 is provided with an opening, and the opening is provided with an adaptable container cover 120, a steam generator 140 is relatively provided on the inner wall of the container body 110, and a heat exchange device 200 is provided on the steam generator 140 so that the outer wall surface 220 is close to the steam generator 140.
  • the container body 110 is cylindrical and vertically upward, mainly used to accommodate the nuclear reactor core 500 and the internal coolant, and is made of stainless steel.
  • the steam generator 140 is arranged between the outer wall 220 of the heat exchange device 200 and the inner wall of the container body 110, that is, the upper part of the descending annular cavity 171, and the height is higher than the core 500 and as high as possible.
  • the steam generator 140 is a normal heat removal system in a nuclear reactor.
  • the steam generator 140 directly exchanges heat with the liquid coolant in the reactor vessel 100, and the high-temperature superheated steam generated drives the turbine and then drives the generator to generate electricity.
  • the isolation valve in the water supply device 400 is not opened; after an accident occurs in a nuclear power plant, the steam generator 140 in the normal heat removal system cannot work.
  • the second cavity 250 in the heat exchange device 200 is used to cooperate with the water supply device 400 to perform heat exchange to discharge the residual heat of the core 500.
  • the steam generator 140 is composed of a spiral heat exchange tube bundle, the interior of the tube bundle is provided with cold water by the main water supply system of the nuclear reactor, and the exterior of the tube bundle is immersed in the coolant of the nuclear reactor.
  • the spiral heat exchange tube is made of high-strength steel alloy material, and can transfer heat in the nuclear reactor coolant to the inner wall of the heat exchange tube, gradually heating the cold water flowing inside it. After absorbing heat, the cold water gradually vaporizes and becomes high-temperature and high-pressure superheated steam, which is discharged from the end of the heat exchange tube and enters the main steam system of the nuclear reactor, driving the turbine, and then driving the generator to generate power.
  • the heat exchange device 200 further includes a coolant pipe 270, one end of the coolant pipe 270 is connected to the water supply device 400, and the other end is connected to the second cavity 250, and the coolant pipe 270 includes an outlet pipe 271 and an inlet pipe 272.
  • the outlet pipe 271 is disposed above the inlet pipe 272 .
  • the outlet pipe 271 and the inlet pipe 272 can be arranged at multiple angles in the radial cross section of the reactor vessel 100, and the angles and numbers of the arrangements can be set according to actual needs.
  • the outlet pipe 271 is arranged above the inlet pipe 272.
  • the outlet pipe 271 is arranged at one end of the heat exchange device 200 away from the core 500, and the inlet pipe 272 is arranged at one end of the heat exchange device 200 close to the bottom of the core 500, so that the cold source water in the second cavity 250 can fully realize heat exchange therein.
  • the outlet pipe 271 and the inlet pipe 272 penetrate the wall of the reactor vessel 100, one end is connected to the second cavity 250 of the heat exchange device 200, and the other end is connected to the water supply device 400 for providing circulating cold source water.
  • the second cavity 250 of the heat exchange device 200 is filled with non-condensable inert gas and will not exchange heat with the liquid coolant in the nuclear reactor.
  • the cold source water in the water supply device 400 enters the inlet pipe 272 due to gravity and enters the bottom of the heat exchange device 200, before the water level rises to the height of the pipe mouth of the outlet pipe 271, the cold source water contacts and exchanges heat with the two side walls and the heat dissipation fins 260 in the second cavity 250. After being heated by the heat of the core 500, the cold source water will be heated and evaporated, and the generated water vapor is discharged from the outlet pipe 271.
  • the core 500 includes a fuel rod bundle and a control rod assembly.
  • the nuclear power of the nuclear reaction of the fuel rods is controlled by controlling the insertion depth of the control rod assembly.
  • the outer side of the core 500 is wrapped with a neutron reflection layer 510.
  • the outer wall surface 220 contacts the coolant in the descending annular cavity 171 for heat exchange, part of the inner wall surface 210 contacts the neutron reflector 510 for heat exchange, and part of the inner wall surface 210 contacts the coolant in the upper chamber 160 for heat exchange, even under the condition of low natural circulation flow of the core 500, the decay heat of the core 500 can still be continuously discharged through the inner wall surface 210 of the heat exchange device 200.
  • the heat generated by the nuclear reaction in the core 500 is transferred to the coolant flowing through the core 500 through heat exchange, and by wrapping a layer of neutron reflector 510 around the core 500 with the function of reflecting neutrons back to the core 500, the loss of fission neutrons to the surrounding environment is reduced, and the burnup depth of the nuclear reactor is increased.
  • the isolation valve is in a closed state, and the second cavity 250 of the heat exchange device 200 is filled with non-condensable inert gas, and does not exchange heat with the coolant in the reactor core 500.
  • the reactor is shut down, and the isolation valve is activated by an automatic signal to start the emergency residual heat removal operation.
  • the water supply device 400 is configured as a pool-type water supply device 400, including a pit 410 for accommodating the reactor vessel 100, and a pipe through hole 130 is provided on the reactor vessel 100, and the outlet pipe 271 and the inlet pipe 272 are both connected to the pit 410 through the pipe through hole 130.
  • the entire reactor vessel 100 is arranged in the sunken pit 410, and the outlet pipe 271 and the inlet pipe 272 of the heat exchange device 200 are directly connected to the heat exchange device 200 through the pipe through hole 130. It is connected to the inner space of the pit 410 , and no additional isolation valve is required between the two, which simplifies the structure of the reactor vessel 100 .
  • the water supply device 400 can directly inject cold source water into the pit 410 .
  • the water supply device may include a cold source water tank 420 and a drain pipe 430.
  • the drain pipe 430 is provided to connect the cold source water tank 420 and the pit 410.
  • a first isolation valve 440 for controlling the on-off is provided on the drain pipe 430.
  • the height of the cold source water tank 420 is higher than the pit 410, and an exhaust window 450 connected to the outside is provided on the upper end surface of the cold source water tank 420.
  • the cold source water tank 420 is set on the upper part of the reactor vessel 100 and the pit 410, and the first isolation valve 440 is set on the drain pipe 430 to control the on and off of the cold source water.
  • the exhaust window 450 on the top of the cold source water tank 420 plays the role of sucking the gas outside the water tank into the water tank to balance the pressure inside the water tank, ensuring that the cooling water in the water tank can be smoothly injected into the pit 410.
  • the diameter of the drain pipe 430 can be appropriately increased, and the volume of the pit 410 below the height of the outlet pipe 271 can be reduced.
  • the working process of nuclear reactor waste heat discharge under this scheme is as follows: when the waste heat of the core 500 cannot be discharged normally, the emergency waste heat signal is sent to trigger the first isolation valve 440 to open, and the cooling water of the cold source water tank 420 is injected into the pit 410 through the drain pipe 430 under the action of gravity. When the water level in the pit 410 rises to submerge the inlet pipe 272, the cooling water begins to enter the second cavity 250 of the heat exchange device 200, thereby contacting and exchanging heat with the heat dissipation fins 260 in the second cavity 250 and the wall surface of the second cavity 250.
  • Part of the cooling water in the second cavity 250 will be heated and evaporated, and the generated water vapor will be discharged from the outlet pipe 271, forming a natural circulation flow channel of cooling water from the cold source water tank 420-pit 410-inlet pipe 272-second cavity 250 of the heat exchange device-outlet pipe 271-pit 410, which continuously brings out the decay heat inside the nuclear reactor.
  • the water supply device 400 is configured as a tubular water supply device 400, including an inlet mother pipe 460, an outlet mother pipe 470 and a cold source water tank 420, multiple outlet pipes 271 and multiple inlet pipes 272 are circumferentially arranged along the outer wall of the heat exchange device 200, multiple inlet pipes 272 merge to form the inlet mother pipe 460, and multiple outlet pipes 271 merge to form the outlet mother pipe 470.
  • the inlet mother pipe 460 and the outlet mother pipe 470 are respectively used to connect the cold source water tank 420 with the second cavity 250, a second isolation valve 480 is provided on the outlet mother pipe 470, and a third isolation valve 490 is provided on the inlet mother pipe 460.
  • the height of the cold source water tank 420 is higher than the reactor vessel 100, and an exhaust window 450 communicating with the outside is provided on the upper end surface of the cold source water tank 420.
  • the cold source water tank 420 is arranged at a position as high as possible on the upper part of the reactor vessel 100, and the outlet mother pipe 470 and the inlet mother pipe are respectively connected to the bottom of the cold source water tank 420, thereby forming a circulating heat exchange loop of cold source water tank 420-inlet mother pipe 460-inlet pipe 272-second cavity 250 of heat exchange device 200-outlet pipe 271-outlet mother pipe 470-cold source water tank 420.
  • the working process of the nuclear reactor residual heat removal under this scheme is as follows: when the residual heat of the core 500 cannot be discharged normally, the emergency residual heat signal is sent to trigger the second isolation valve 480 and the third isolation valve 490 to open at the same time, and the cooling water in the cold source water tank 420 is turned off. Under the action of gravity, it is injected into the inlet main pipe 460, and then enters the inlet pipe 272 of the heat exchange device 200, pushing away the non-condensable gas originally in the second cavity 250 of the heat exchange, and the non-condensable gas enters the outlet pipe 271, then reaches the outlet main pipe, and finally enters the cold source water tank 420.
  • the cooling water is in contact with the heat dissipation fins 260 and the wall of the second cavity 250 for heat exchange, the temperature gradually rises, and the density gradually decreases, so that the gravity pressure head of the water inlet main pipe is greater than the gravity pressure head of the water outlet main pipe, forming a gravity difference driving force, and then forming a circulation heat exchange loop of cold source water tank 420-inlet main pipe-inlet pipe 272-heat exchange device 200-outlet pipe 271-outlet main pipe-cold source water tank 420, which continuously brings out the decay heat of the reactor. Under this scheme, the water temperature in the cold source water tank 420 gradually rises, and the steam generated after boiling will be discharged to the external environment through the exhaust window 450 on the top of the water tank.
  • the cold source water tank 420 of the above two schemes can be arranged into various irregular shapes such as square and cylindrical according to the characteristics of the nuclear reactor building, and can be made of stainless steel or concrete with stainless steel lining. A large amount of deionized water is filled in the cold source water tank 420, and the initial water volume can be set according to the power characteristics of the reactor.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Chemical & Material Sciences (AREA)
  • Chemical Kinetics & Catalysis (AREA)
  • Heat-Exchange Devices With Radiators And Conduit Assemblies (AREA)

Abstract

一种核反应堆，包括：反应堆容器(100)、热交换装置(200)、驱动装置(300)和供水装置(400)。其中，反应堆容器(100)内有液体冷却剂和气腔(150)，热交换装置(200)设置于反应堆容器(100)，热交换装置(200)包括外层壁面(220)和内层壁面(210)，其中内层壁面(210)形成容纳堆芯(500)的第一空腔(240)，外层壁面(220)和内层壁面(210)共同限定出第二空腔(250)，驱动装置(300)一端与气腔(150)连接，另一端靠近堆芯(500)设置，供水装置(400)与第二空腔(250)连通。该核反应堆可在发生事故工况时，排出反应堆一次侧和堆芯(500)的热量，避免堆芯(500)进一步升温导致燃料损坏。

Description

一种核反应堆
相关申请
本申请要求2022年12月22日申请的，申请号为2022116562535，名称为“一种核反应堆”的中国专利申请的优先权，在此将其全文引入作为参考。
技术领域
本申请涉及核电设备技术领域，特别是涉及一种核反应堆。
背景技术
为了满足特定应用场景(如偏远矿区、极地科考、应急救灾等)的供电供热需求，国际上开展了具有固有安全、系统简单、具有工厂全预制和快速拼装能力的小型移动核反应堆研究，通常为了便于反应堆的安装、固定和运输，小型移动核反应堆的安全壳设置为圆筒状，由此产生的问题是反应堆堆芯和换热器之间的高度差较小，循环能力较弱。
当反应堆发生意外紧急情况(事故工况)时，反应堆正常的热量排出系统无法正常排出反应堆一次侧和堆芯的热量，反应堆的一次侧将会出现高温的状态，为了避免堆芯进一步升温导致燃料损坏，反应堆内产生的衰变热必须排出，防止出现升压失控，此时就需要有核反应堆能够发挥作用，带走反应堆一回路和堆芯热量，将意外发生后核反应堆安全壳的内部压力在尽可能短的时间内恢复到正常状态。
相关技术中的核反应堆大多需借助电源动力和类似主泵的能动部件，同时由于常规反应堆的一回路阻力较大，为了驱动冷却剂循环，往往需要配置多个大体积的主泵，采用能动部件使整个系统的复杂化，提高了造价成本，也无法避免失效造成的系统可靠性较低的问题。
发明内容
基于此，有必要针对相关技术中核反应堆由于采用能动部件进行循环而造成的系统设计复杂化，造价提高，以及可靠性较低的问题，提供一种不依赖能动部件的非能动热量排出系统，提高反应堆安全性，同时简化系统设计方案，降低整体系统造价的核反应堆。
一种核反应堆，包括：
反应堆容器，所述反应堆容器内设置有液体冷却剂，所述反应堆容器内位于所述液体 冷却剂上方的部分形成气腔；
热交换装置，所述热交换装置设置于所述反应堆容器内且位于所述液体冷却剂的液面以下，所述热交换装置包括内层壁面和外层壁面，所述内层壁面围绕形成用于容纳堆芯的第一空腔，所述外层壁面和所述内层壁面共同限定出密闭的第二空腔；
驱动装置，所述驱动装置设置于所述反应堆容器且位置高于所述液体冷却剂的液面，包括有抽吸端和回注端，所述抽吸端与所述气腔连通，所述回注端伸入所述第一空腔且靠近堆芯设置；
供水装置，所述供水装置与所述第二空腔连通，用于提供所述热交换装置所需的循环冷源水。
在其中一个实施例中，所述第二空腔相对的两侧壁面上设置有多个散热翅片，多个所述散热翅片沿所述第二空腔两侧壁面周向间隔设置，所述散热翅片沿所述第二空腔侧壁竖直方向延伸设置。
在其中一个实施例中，所述反应堆容器内部设置多个支撑件，包括设置于反应堆容器内侧壁的第一支撑件和设置于反应堆容器内底部的第二支撑件，所述热交换装置配合所述支撑件悬空设置于所述反应堆容器。
在其中一个实施例中，所述液体冷却剂为熔融金属冷却剂，所述气腔内填充有惰性气体。
在其中一个实施例中，所述驱动装置包括气体压缩泵，所述气体压缩泵一端设置有与所述气腔连通的抽吸管，另一端设置有回注管，靠近所述堆芯设置有环形喷气管，沿所述环形喷气管周向设置多个朝上的喷嘴，所述回注管一端与气体压缩泵连接，另一端与环形喷气管连接。
在其中一个实施例中，所述反应堆容器包括容器本体和容器上盖，所述容器本体的上端面设置有开口，所述开口上设置有适配的所述容器上盖，所述容器本体内侧壁上相对设置有蒸汽发生器，所述热交换装置设置于所述蒸汽发生器，以使所述外层壁面贴近所述蒸汽发生器。
在其中一个实施例中，所述热交换装置还包括冷却剂管道，所述冷却剂管道一端与所述供水装置连通，另一端与所述第二空腔连通，所述冷却剂管道包括出口管道和入口管道，所述出口管道设置于所述入口管道上方。
在其中一个实施例中，所述堆芯包括燃料棒束和控制棒组件，通过控制所述控制棒组件插入的深度，达到控制燃料棒发生核反应的核功率，所述堆芯外侧包裹有中子反射层。
在其中一个实施例中，所述供水装置设置为池式供水装置，包括用于容纳所述反应堆 容器的地坑，所述反应堆容器上设置有管道通孔，所述出口管道和所述入口管道均通过所述管道通孔与所述地坑连通。
在其中一个实施例中，所述供水装置设置为管式供水装置，包括入口母管道、出口母管道和冷源水箱，多个所述出口管道和多个所述入口管道沿所述热交换装置外侧壁面上周向设置，多个所述入口管道汇合形成所述入口母管道，多个所述出口管道汇合形成所述出口母管道，所述入口母管道、所述出口母管道分别用于将所述冷源水箱与所述第二空腔连通，所述出口母管道上设置有第二隔离阀，所述入口母管道上设置有第三隔离阀，所述冷源水箱的高度高于所述反应堆容器，且所述冷源水箱上端面设置有与外界连通的排气窗口。
上述说明仅是本申请技术方案的概述，为了能够更清楚了解本申请的技术手段，而可依照说明书的内容予以实施，并且为了让本申请的上述和其它目的、特征和优点能够更明显易懂，以下特举本申请的具体实施方式。
附图说明
为了更清楚地说明本申请实施例或传统技术中的技术方案，下面将对实施例或传统技术描述中所需要使用的附图作简单地介绍，显而易见地，下面描述中的附图仅仅是本申请的实施例，对于本领域普通技术人员来讲，在不付出创造性劳动的前提下，还可以根据公开的附图获得其他的附图。
图1为核反应堆的正视剖面图(不包括供水装置)；
图2为核反应堆的反应堆容器的内部横截面示意图；
图3为核反应堆的热交换装置正视剖面图；
图4为核反应堆的热交换装置俯视剖面图；
图5为核反应堆采用池式供水装置的结构示意图；
图6为核反应堆采用管式供水装置的结构示意图。
附图标记说明：
100、反应堆容器；110、容器本体；120、容器上盖；130、管道通孔；140、蒸汽发生器；150、气腔；160、上腔室；170、下腔室；171、下降环腔；200、热交换装置；210、内层壁面；220、外层壁面；230、连接壁面；240、第一空腔；250、第二空腔；260、散热翅片；270、冷却剂管道；271、出口管道；272、入口管道；300、驱动装置；310、气体压缩泵；320、抽吸管；330、回注管；331、第一管体；332、第二管体；340、环形喷气管；341、喷嘴；400、供水装置；410、地坑；420、冷源水箱；430、排水管；440、第一隔离阀；450、排气窗口；460、入口母管道；470、出口母管道；480、第二隔离阀；490、 第三隔离阀；500、堆芯；510、中子反射层。
具体实施方式
下面将结合本申请实施例中的附图，对本申请实施例中的技术方案进行清楚、完整地描述，显然，所描述的实施例仅仅是本申请一部分实施例，而不是全部的实施例。基于本申请中的实施例，本领域普通技术人员在没有做出创造性劳动前提下所获得的所有其他实施例，都属于本申请保护的范围。
下面将结合附图对本申请技术方案的实施例进行详细的描述。以下实施例仅用于更加清楚地说明本申请的技术方案，因此只作为示例，而不能以此来限制本申请的保护范围。
除非另有定义，本文所使用的所有的技术和科学术语与属于本申请的技术领域的技术人员通常理解的含义相同；本文中所使用的术语只是为了描述具体的实施例的目的，不是旨在于限制本申请；本申请的说明书和权利要求书及上述附图说明中的术语“包括”和“具有”以及它们的任何变形，意图在于覆盖不排他的包含。
在本申请实施例的描述中，技术术语“第一”“第二”等仅用于区别不同对象，而不能理解为指示或暗示相对重要性或者隐含指明所指示的技术特征的数量、特定顺序或主次关系。在本申请实施例的描述中，“多个”的含义是两个以上，除非另有明确具体的限定。
在本文中提及“实施例”意味着，结合实施例描述的特定特征、结构或特性可以包含在本申请的至少一个实施例中。在说明书中的各个位置出现该短语并不一定均是指相同的实施例，也不是与其它实施例互斥的独立的或备选的实施例。本领域技术人员显式地和隐式地理解的是，本文所描述的实施例可以与其它实施例相结合。
在本申请实施例的描述中，术语“和/或”仅仅是一种描述关联对象的关联关系，表示可以存在三种关系，例如A和/或B，可以表示：单独存在A，同时存在A和B，单独存在B这三种情况。另外，本文中字符“/”，一般表示前后关联对象是一种“或”的关系。
在本申请实施例的描述中，术语“多个”指的是两个以上(包括两个)，同理，“多组”指的是两组以上(包括两组)，“多片”指的是两片以上(包括两片)。
在本申请实施例的描述中，技术术语“中心”“纵向”“横向”“长度”“宽度”“厚度”“上”“下”“前”“后”“左”“右”“竖直”“水平”“顶”“底”“内”“外”“顺时针”“逆时针”“轴向”“径向”“周向”等指示的方位或位置关系为基于附图所示的方位或位置关系，仅是为了便于描述本申请实施例和简化描述，而不是指示或暗示所指的装置或元件必须具有特定的方位、以特定的方位构造和操作，因此不能理解为对本申请实施例的限制。
在本申请实施例的描述中，除非另有明确的规定和限定，技术术语“安装”“相连”“连接”“固定”等术语应做广义理解，例如，可以是固定连接，也可以是可拆卸连接，或成一体；也可以是机械连接，也可以是电连接；可以是直接相连，也可以通过中间媒介间接相连，可以是两个元件内部的连通或两个元件的相互作用关系。对于本领域的普通技术人员而言，可以根据具体情况理解上述术语在本申请实施例中的具体含义。
相关技术中核反应堆往往采用回路换热器的形式，以形成自然循环的动力，动力的大小取决于热源和冷源的相对高位差，如果高差太小则无法形成自然循环。采用回路换热器需要安装换热器来作为传热界面，对于小型一体化的反应堆，没有多余的空间安装换热器，并且回路换热器采用的是直径较小，壁厚较薄的细管，在高温高压条件下容易破裂，可靠性不高。
参阅图1和图2，本申请提供了一种核反应堆，包括：反应堆容器100、热交换装置200、驱动装置300和供水装置400，其中反应堆容器100，反应堆容器100内设置有液体冷却剂，反应堆容器100内位于液体冷却剂上方的部分形成气腔150，热交换装置200，热交换装置200设置于反应堆容器100内且位于液体冷却剂的液面以下，热交换装置200包括内层壁面210和外层壁面220，内层壁面210围绕形成用于容纳堆芯500的第一空腔240，外层壁面220和内层壁面210共同限定出密闭的第二空腔250，驱动装置300，驱动装置300设置于反应堆容器100且位置高于液体冷却剂液面，包括有抽吸端和回注端，抽吸端与气腔150连通，回注端伸入第一空腔240且靠近堆芯500设置，供水装置400，供水装置400与第二空腔250连通，用于提供热交换装置200所需的循环冷源水。
该核反应堆包括反应堆容器100、热交换装置200、驱动装置300和供水装置400，其中反应堆容器100内设置有液体冷却剂，位于液体冷却剂上方的部分形成气腔150，通过在反应堆容器100中液体冷却剂的液面以下设置有热交换装置200，热交换装置200的内层壁面210围成的第一空腔240用于容纳堆芯500，外层壁面220和内层壁面210共同限定出密闭的第二空腔250与供水装置400连通。
当核电厂正常运行时，反应堆堆芯500功率较大，通过驱动装置300抽吸端在气腔150中抽取气体再经回注端注入堆芯500顶部的液体冷却剂中，注入气体使得堆芯500顶部以上的液体冷却剂平均密度小于同等高度下蒸汽发生器140附近的冷却剂平均密度，进而产生重力差，驱动液体冷却剂在反应堆容器100内部循环流动，以形成从第一空腔堆芯500上方出口-上腔室160-下降环腔171-下腔室170-堆芯500入口-堆芯500-堆芯500上方出口的循环路线，液体冷却剂在反应堆容器100中自上而下流动传递堆芯500产生的热量至布置在下降环腔171中的蒸汽发生器140。当核电厂发生事故时，驱动装置300停止运作， 堆芯500配置的热量排出系统无法正常工作，供水装置400的中的循环冷源水通过在第二空腔250中循环，不断与堆芯500进行热交换，用于排出堆芯500余热。本反应堆无需设置大体积主泵，降低了整个系统的复杂度和造价成本，避免了因热量排出系统失效造成系统可靠性低的问题。
示例性的，外层壁面220设置于内层壁面210外侧，热交换装置200还包括用以限定出第二空腔250的连接壁面230，连接壁面230设置于外层壁面220和内侧壁面两端面之间，形成了四周密封的环形空腔，通过热交换装置200的内、外层壁面220与第一空腔240堆芯500的冷却剂接触，可将冷却剂的热量通过热传导的方式传导至第二空腔250内壁面以及焊接在第二空腔250内壁面的散热翅片260上。本实施例中的热交换装置200采用不锈钢材料制造而成，用于承担核反应堆停堆状态下的余热导出功能，同时将堆芯500和外侧间隔为两个不同的循环通道，使该核反应堆无需另外配置堆芯500吊篮，节省设备造价和空间。
本申请可选的实施例中，进一步参考图3和图4，第二空腔250相对的两侧壁面上设置有多个散热翅片260，多个散热翅片260沿第二空腔250两侧壁面周向间隔设置，散热翅片260沿第二空腔250侧壁竖直方向延伸设置。本实施例中设置散热翅片260为薄片结构，用于增大与冷源水接触的换热面积，提高换热效率。
示例性的，反应堆容器100内部设置多个支撑件(图中未示出)，包括设置于反应堆容器100内侧壁的第一支撑件和设置于反应堆容器100内底部的第二支撑件，热交换装置200配合支撑件悬空设置于反应堆容器100，实现更好的热交换效果。
示例性的，通过多个支撑件将热交换装置200悬空设置于反应堆容器100内部，此时反应堆容器100中堆芯500上方，液体冷却剂液面以下形成有上腔室160，反应堆容器100中堆芯500下方形成有下腔室170，下腔室170还包括热交换装置200外层壁面220与反应堆容器100内壁面之间的下降环腔171。
示例性的，液体冷却剂为熔融金属冷却剂，气腔150内填充有惰性气体。具体的，核反应堆的熔融金属冷却剂可采用多种不同类型的液体，包括液态钠，铅、铅铋合金等，由于液态金属冷却剂的密度大，导热性能好，因此在该核反应堆中采用熔融金属冷却剂可较大程度上增加反应堆的安全性能。此外在熔融金属冷却剂液面的气腔150中充有惰性气体，用以控制熔融金属冷却剂的压力，保持在一定安全范围内。另外，一回路采用熔融金属作为冷却剂，不需要太多的冷热源高位差，能够节省布置空间。
本申请可选的实施例中，进一步参考图5和图6，驱动装置300包括气体压缩泵310，气体压缩泵310一端设置有与气腔150连通的抽吸管320，另一端设置有回注管330，靠 近堆芯500设置有环形喷气管340，沿环形喷气管340周向设置多个朝上的喷嘴341，回注管330一端与气体压缩泵310连接，另一端与环形喷气管340连接。
示例性的，本实施例中为了避免抽吸管320和回注管330发生破口时，造成核反应堆内的液体冷却剂流失，通过将气体压缩泵310设置在高于反应堆容器100内部液体冷却剂液面的高度，即抽吸管320和回注管330两条管线穿设于反应堆容器100壁面的位置均高于内部冷却剂的液面位置。回注管330包括穿设于反应堆容器100的第一管体331和转弯向下沿热交换装置200内层壁面210向下延伸的第二管体332，第二管体332和环形喷气管340连接，多个间隔分布在环形喷气管340的喷嘴341能够更加均匀的排出气体。使用配置更简单的气体压缩泵310代替主泵，通过调节气体压缩泵310的流量改变重力差，进而改变堆芯500的冷却剂流量，使整个核反应堆的流道简单，减小产生的阻力，提高自然循环能力。
本申请可选的实施例中，进一步参考图1，反应堆容器100包括容器本体110和容器上盖120，容器本体110上端面设置有开口，开口上设置有适配的容器上盖120，容器本体110内侧壁上相对设置有蒸汽发生器140，热交换装置200设置于蒸汽发生器140，以使外层壁面220贴近蒸汽发生器140。
本实施例中设置容器本体110为圆筒状竖直向上，主要用于容纳其内部核反应堆堆芯500和内部冷却剂，由不锈钢材料制作而成。蒸汽发生器140设置于热交换装置200外层壁面220与容器本体110内壁面之间，即下降环腔171的上部，高度高于堆芯500，且尽可能高。
示例性的，蒸汽发生器140属核反应堆中正常的热量排出系统，正常工作时通过蒸汽发生器140直接与反应堆容器100中的液体冷却剂进行热交换，产生的高温过热蒸汽推动涡轮机进而带动发电机发电。此种情况下供水装置400中的隔离阀不开启；当核电厂发生事故之后，正常的热量排出系统中的蒸汽发生器140不能工作，此种情况下应用热交换装置200中的第二空腔250配合供水装置400进行热交换以排出堆芯500余热。
示例性的，该蒸汽发生器140由螺旋换热管束组成，管束内部由核反应堆的主给水系统提供冷水，管束外部浸没在核反应堆的冷却剂中，螺旋换热管由高强度的钢合金材料制作而成，可将核反应堆冷却剂中的热量传递至换热管内壁面，逐渐加热在其内部流动的冷水，冷水吸热后逐渐汽化，变成高温高压的过热蒸汽从换热管末端排出，进入核反应堆的主蒸汽系统，推动涡轮机，进而带动发电机产生电源。
示例性的，热交换装置200还包括冷却剂管道270，冷却剂管道270一端与供水装置400连通，另一端与第二空腔250连通，冷却剂管道270包括出口管道271和入口管道272， 出口管道271设置于入口管道272上方。
出口管道271、入口管道272在反应堆容器100的径向截面可多角度进行布置，布置的角度和数量可以根据实际需要设置。本实施例中将出口管道271设置于入口管道272上方，具体的，为了保证热交换装置200中第二空腔250的循环效果，将出口管道271设置于热交换装置200远离堆芯500的一端，将入口管道272设置于热交换装置200贴近堆芯500底部一端，使第二空腔250中的冷源水在其中充分实现热交换。
出口管道271和入口管道272穿透反应堆容器100壁面后一端与热交换装置200的第二空腔250连通，另一端与提供循环冷源水的供水装置400连通，当冷源水未进入热交换装置200中时，热交换装置200第二空腔250中充满了不可凝结的惰性气体，不会与核反应堆内的液体冷却剂交换热量，当供水装置400中的冷源水由于重力的驱动进入入口管道272，进入热交换装置200底部，在水位上升至出口管道271的管口高度之前，冷源水与第二空腔250中的两侧壁面和散热翅片260接触换热，受到堆芯500热量的加热后，冷源水会被加热蒸发，从而生成的水蒸汽从出口管道271排出。
示例性的，堆芯500包括燃料棒束和控制棒组件，通过控制控制棒组件插入的深度，达到控制燃料棒发生核反应的核功率，堆芯500外侧包裹有中子反射层510。
示例性的，由于热交换装置200浸没在反应堆容器100的液体冷却剂中，外层壁面220与下降环腔171中的冷却剂接触换热，内层壁面210部分位置与中子反射层510接触换热，部分位置与上腔室160内的冷却剂接触换热，即使在堆芯500自然循环流量较低的工况下，依然可通过热交换装置200内层壁面210源源不断地导出堆芯500衰变热。从而将堆芯500发生核反应产生的热量通过热交换传递至流过堆芯500内部的冷却剂，通过在堆芯500周围包裹一层具备将中子反射回堆芯500功能的中子反射层510，减少裂变中子向周围环境的损失，提高核反应堆的燃耗深度。
示例性的，当核反应堆正常运行时，热交换装置200的应急余热排出功能不启用，隔离阀处于关闭状态，热交换装置200的第二空腔250中充满不可凝结惰性气体，不与反应堆堆芯500内的冷却剂交换热量。当核反应堆反生一、二次侧事故后，进行停堆操作，并由自动信号触发隔离阀启动，开始进行应急余热排出操作。
本申请可选的实施例中，进一步参考图5，供水装置400设置为池式供水装置400，包括用于容纳反应堆容器100的地坑410，反应堆容器100上设置有管道通孔130，出口管道271和入口管道272均通过管道通孔130与地坑410连通。
示例性的，当核反应堆采用池式供水装置400时，将整个反应堆容器100布置在内陷的地坑410中，热交换装置200的出口管道271和入口管道272均通过管道通孔130直接 与地坑410内部空间连通，二者之间无需额外连接隔离阀，简化了反应堆容器100结构，供水装置400可直接向地坑410中注入冷源水。
示例性的，供水装置可以包括冷源水箱420和排水管430，通过设置排水管430用于将冷源水箱420和地坑410连通，排水管430上设置有控制通断的第一隔离阀440，冷源水箱420的高度高于地坑410，且冷源水箱420上端面设置有与外界连通的排气窗口450。
为了满足高度差，将冷源水箱420设置在反应堆容器100和地坑410的上部，在排水管430设置第一隔离阀440用于控制冷源水的通断。冷源水箱420顶部的排气窗口450起到吸取水箱外部气体进入水箱内部平衡压力的作用，确保水箱内的冷却水可顺利地注入地坑410中。为了使地坑410内的水位尽快上升至淹没入口管道272，可适当增大排水管430的管径，同时缩小出口管道271高度以下部分地坑410的体积。
此方案下的核反应堆余热排出的工作过程为：当堆芯500余热无法正常排出时，应急余热信号发出触发第一隔离阀440开启，冷源水箱420的冷却水在冷源水箱420内的冷却水在重力作用下通过排水管430被注入地坑410中，地坑410内的水位上升至淹没入口管道272时，冷却水开始进入热交换装置200第二空腔250内部，从而与第二空腔250腔内的散热翅片260及第二空腔250的壁面接触换热，第二空腔250腔内的部分冷却水会被加热蒸发，生成的水蒸汽从出口管道271排出，形成由冷源水箱420-地坑410-入口管道272-热交换装置第二空腔250-出口管道271-地坑410的冷却水自然循环流道，源源不断地带出核反应堆内部的衰变热。
本申请可选的实施例中，进一步参考图6，供水装置400设置为管式供水装置400，包括入口母管道460、出口母管道470和冷源水箱420，多个出口管道271和多个入口管道272沿热交换装置200外侧壁面上周向设置，多个入口管道272汇合形成入口母管道460，多个出口管道271汇合形成出口母管道470，入口母管道460、出口母管道470分别用于将冷源水箱420与第二空腔250连通，出口母管道470上设置有第二隔离阀480，入口母管道460上设置有第三隔离阀490，冷源水箱420的高度高于反应堆容器100，且冷源水箱420上端面设置有与外界连通的排气窗口450。
示例性的，当核反应堆采用管式供水装置400时，冷源水箱420布置在反应堆容器100的上部高度尽可能高的位置，出口母管道470和入口母管分别与冷源水箱420底部连接，从而形成冷源水箱420-入口母管道460-入口管道272-热交换装置200第二空腔250-出口管道271-出口母管道470-冷源水箱420的循环换热回路。
此方案下的核反应堆余热排出的工作过程为：当堆芯500余热无法正常排出时，应急余热信号发出触发第二隔离阀480和第三隔离阀490同时开启，冷源水箱420内的冷却水 在重力作用下被注入入口母管道460中，再进入热交换装置200的入口管道272，推走原本在热交换第二空腔250内的不可凝结气体，不可凝结气体进入出口管道271，然后到达出口母管，最后进入冷源水箱420。
热交换装置200内充满冷却水后，冷却水被与散热翅片260及第二空腔250的壁面接触换热，温度逐渐上升，密度逐渐变小，从而使得进水母管道的重力压头大于出水母管道的重力压头，形成重力差驱动力，进而形成冷源水箱420-入口母管-入口管道272-热交换装置200-出口管道271-出口母管-冷源水箱420的循环换热回路，源源不断地带出反应堆的衰变热。此方案下，冷源水箱420内的水温逐渐上升，沸腾后产生的蒸汽将通过水箱顶部的排气窗口450排放至外部环境。
上述两种方案的冷源水箱420可根据核反应堆厂房的特点布置成方形、圆柱形等各种不规则形状，具体可由不锈钢材料或混凝土内加不锈钢衬里制作而成。冷源水箱420内部装有大量去离子水，初始水体积可根据反应堆功率特点而设定。
本说明书中各个实施例采用递进的方式描述，每个实施例重点说明的都是与其他实施例的不同之处，各个实施例之间相同相似部分互相参见即可。
对所公开的实施例的上述说明，使本领域专业技术人员能够实现或使用本申请。对这些实施例的多种修改对本领域的专业技术人员来说将是显而易见的，本文中所定义的一般原理可以在不脱离本申请的精神或范围的情况下，在其它实施例中实现。因此，本申请将不会被限制于本文所示的这些实施例，而是要符合与本文所公开的原理和新颖特点相一致的最宽的范围。
以上所述实施例的各技术特征可以进行任意的组合，为使描述简洁，未对上述实施例中的各个技术特征所有可能的组合都进行描述，然而，只要这些技术特征的组合不存在矛盾，都应当认为是本说明书记载的范围。
以上所述实施例仅表达了本申请的几种实施方式，其描述较为具体和详细，但并不能因此而理解为对申请专利范围的限制。应当指出的是，对于本领域的普通技术人员来说，在不脱离本申请构思的前提下，还可以做出若干变形和改进，这些都属于本申请的保护范围。因此，本申请专利的保护范围应以所附权利要求为准。

Claims (10)

  1. 一种核反应堆，其特征在于，包括：
    反应堆容器(100)，所述反应堆容器(100)内设置有液体冷却剂，所述反应堆容器(100)内位于所述液体冷却剂上方的部分形成气腔(150)；
    热交换装置(200)，所述热交换装置(200)设置于所述反应堆容器(100)内且位于所述液体冷却剂的液面以下，所述热交换装置(200)包括内层壁面(210)和外层壁面(220)，所述内层壁面(210)围绕形成用于容纳堆芯(500)的第一空腔(240)，所述外层壁面(220)和所述内层壁面(210)共同限定出密闭的第二空腔(250)；
    驱动装置(300)，所述驱动装置(300)设置于所述反应堆容器(100)且位置高于所述液体冷却剂的液面，包括有抽吸端和回注端，所述抽吸端与所述气腔(150)连通，所述回注端伸入所述第一空腔(240)且靠近所述堆芯(500)设置；
    供水装置(400)，所述供水装置(400)与所述第二空腔(250)连通，用于提供所述热交换装置(200)所需的循环冷源水。
  2. 根据权利要求1所述的核反应堆，其特征在于，所述第二空腔(250)相对的两侧壁面上设置有多个散热翅片(260)，多个所述散热翅片(260)沿所述第二空腔(250)两侧壁面周向间隔设置，所述散热翅片(260)沿所述第二空腔(250)侧壁竖直方向延伸设置。
  3. 根据权利要求1所述的核反应堆，其特征在于，所述反应堆容器(100)内部设置多个支撑件，包括设置于所述反应堆容器(100)内侧壁的第一支撑件和设置于所述反应堆容器(100)内底部的第二支撑件，所述热交换装置(200)配合所述支撑件悬空设置于所述反应堆容器(100)。
  4. 根据权利要求1所述的核反应堆，其特征在于，所述液体冷却剂为熔融金属冷却剂，所述气腔(150)内填充有惰性气体。
  5. 根据权利要求1所述的核反应堆，其特征在于，所述驱动装置(300)包括气体压缩泵(310)，所述气体压缩泵(310)一端设置有与所述气腔(150)连通的抽吸管(320)，另一端设置有回注管(330)，靠近所述堆芯(500)设置有环形喷气管(340)，沿所述环形喷气管(340)周向设置多个朝上的喷嘴(341)，所述回注管(330)一端与所述气体压缩泵(310)连接，另一端与所述环形喷气管(340)连接。
  6. 根据权利要求1所述的核反应堆，其特征在于，所述反应堆容器(100)包括容器本体(110)和容器上盖(120)，所述容器本体(110)的上端面设置有开口，所述开口上设 置有适配的所述容器上盖(120)，所述容器本体(110)内侧壁上相对设置有蒸汽发生器(140)，所述热交换装置(200)设置于所述蒸汽发生器(140)，以使所述外层壁面(220)贴近所述蒸汽发生器(140)。
  7. 根据权利要求1所述的核反应堆，其特征在于，所述热交换装置(200)还包括冷却剂管道(270)，所述冷却剂管道(270)一端与所述供水装置(400)连通，另一端与所述第二空腔(250)连通，所述冷却剂管道(270)包括出口管道(271)和入口管道(272)，所述出口管道(271)设置于所述入口管道(272)上方。
  8. 根据权利要求1所述的核反应堆，其特征在于，所述堆芯(500)包括燃料棒束和控制棒组件，通过控制所述控制棒组件插入的深度，达到控制燃料棒发生核反应的核功率，所述堆芯(500)外侧包裹有中子反射层(510)。
  9. 根据权利要求7所述的核反应堆，其特征在于，所述供水装置(400)设置为池式供水装置(400)，包括用于容纳所述反应堆容器(100)的地坑(410)，所述反应堆容器(100)上设置有管道通孔(130)，所述出口管道(271)和所述入口管道(272)均通过所述管道通孔(130)与所述地坑(410)连通。
  10. 根据权利要求7所述的核反应堆，其特征在于，所述供水装置(400)设置为管式供水装置(400)，包括入口母管道(460)、出口母管道(470)和冷源水箱(420)，多个所述出口管道(271)和多个所述入口管道(272)沿所述热交换装置(200)外侧壁面上周向设置，多个所述入口管道(272)汇合形成所述入口母管道(460)，多个所述出口管道(271)汇合形成所述出口母管道(470)，所述入口母管道(460)、所述出口母管道(470)分别用于将所述冷源水箱(420)与所述第二空腔(250)连通，所述出口母管道(470)上设置有第二隔离阀(480)，所述入口母管道(460)上设置有第三隔离阀(490)，所述冷源水箱(420)的高度高于所述反应堆容器(100)，且所述冷源水箱(420)上端面设置有与外界连通的排气窗口(450)。
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