WO2016127527A1 - 热交换系统和核反应堆系统 - Google Patents

热交换系统和核反应堆系统 Download PDF

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Publication number
WO2016127527A1
WO2016127527A1 PCT/CN2015/080653 CN2015080653W WO2016127527A1 WO 2016127527 A1 WO2016127527 A1 WO 2016127527A1 CN 2015080653 W CN2015080653 W CN 2015080653W WO 2016127527 A1 WO2016127527 A1 WO 2016127527A1
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Prior art keywords
steam
heat exchange
nuclear reactor
heat
circuit
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PCT/CN2015/080653
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English (en)
French (fr)
Inventor
詹文龙
杨磊
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中国科学院近代物理研究所
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Priority to US15/128,007 priority Critical patent/US20170098483A1/en
Publication of WO2016127527A1 publication Critical patent/WO2016127527A1/zh

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    • 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
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B1/00—Methods of steam generation characterised by form of heating method
    • F22B1/02—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B1/00—Methods of steam generation characterised by form of heating method
    • F22B1/02—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers
    • F22B1/023—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers with heating tubes, for nuclear reactors as far as they are not classified, according to a specified heating fluid, in another group
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B1/00—Methods of steam generation characterised by form of heating method
    • F22B1/02—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers
    • F22B1/16—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being hot liquid or hot vapour, e.g. waste liquid, waste vapour
    • F22B1/162—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being hot liquid or hot vapour, e.g. waste liquid, waste vapour in combination with a nuclear installation
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F22—STEAM GENERATION
    • F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
    • F22B37/00—Component parts or details of steam boilers
    • F22B37/02—Component parts or details of steam boilers applicable to more than one kind or type of steam boiler
    • F22B37/26—Steam-separating arrangements
    • F22B37/268—Steam-separating arrangements specially adapted for steam generators of nuclear power plants
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactor types
    • G21C1/04—Thermal reactors ; Epithermal reactors
    • G21C1/06—Heterogeneous reactors, i.e. in which fuel and moderator are separated
    • G21C1/08—Heterogeneous reactors, i.e. in which fuel and moderator are separated moderator being highly pressurised, e.g. boiling water reactor, integral super-heat reactor, pressurised water reactor
    • G21C1/084—Boiling water reactors
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactor types
    • G21C1/30—Subcritical reactors ; Experimental reactors other than swimming-pool reactors or zero-energy 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
    • G21C15/16—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants comprising means for separating liquid and steam
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G21C19/42—Reprocessing of irradiated fuel
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • G21D1/006—Details of nuclear power plant primary side of steam generators
    • 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
    • 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 invention relates to a heat exchange system and a nuclear reactor system.
  • Nuclear reactor systems typically use liquid metal as the cooling medium.
  • a heat exchange system comprising: a heating device; a heat consuming device, the heat consuming device and the heating device are connected by a pipe to form a circuit; and steam, which is in a wet steam before being supplied to the heat source The state is supplied to the heat consuming device by heat exchange with the heating device to become dry steam or superheated steam.
  • the heat exchange system further includes: a steam-water separation device disposed in the circuit downstream of the steam outlet of the heating device for separating the output from the heating device Liquid water in the steam.
  • the heat exchange system further comprises: a humidity control device disposed in the circuit upstream of a steam inlet of the heating device for controlling the humidity of the steam.
  • the heat exchange system further comprises: a temperature control device disposed in the circuit upstream of the humidity control device for controlling the temperature of the steam.
  • the heat exchange system further comprises: a pressure control device disposed in the circuit upstream of the temperature control device for controlling the pressure of the steam.
  • the heat consuming device is a heat exchanger or a power generation system.
  • the steam is formed from heavy water.
  • a steam inlet of the heating device is disposed on a lower side of the heating device, and a steam outlet of the heating device is disposed on an upper side of the heating device, and the steam-water separation device is disposed at Above the steam outlet.
  • a nuclear reactor system comprising: a nuclear reactor; a heat consuming device, a heat consuming device and a nuclear reactor connected by a pipeline to form a loop; and steam, which is wet steam before being supplied to the nuclear reactor State, and supply heat to the consumer after heat exchange with the nuclear reactor into dry steam or superheated steam.
  • the nuclear reactor system further includes: a steam-water separation device disposed in the circuit downstream of a steam outlet of the nuclear reactor for separating steam output from the nuclear reactor Liquid water.
  • water vapor as a cooling medium has the advantages of large heat capacity, low pressure system, non-corrosive, off-line processing, and the like.
  • This kind of water vapor cooling medium cooled fission reactor can operate safely and reliably at high power density.
  • FIG. 1 is a schematic diagram of a nuclear energy system in accordance with an embodiment of the present invention.
  • Figure 2 is a schematic illustration of a nuclear reactor system in accordance with a first embodiment of the present invention
  • Figure 3 is a schematic illustration of a nuclear reactor system in accordance with a second embodiment of the present invention.
  • FIG. 4 is a schematic diagram of a reactor in accordance with an embodiment of the present invention.
  • Figure 5 is a schematic illustration of a fuel cycle for production capacity in accordance with an embodiment of the present invention.
  • FIG. 1 shows a schematic diagram of a nuclear energy system in accordance with an embodiment of the present invention.
  • a nuclear energy system according to an exemplary embodiment of the present invention includes a nuclear reactor system 100 and a fuel circulation system 200.
  • the nuclear energy system can be a fast neutron nuclear energy system.
  • Nuclear reactor system 100 can be a fast neutron nuclear reactor system.
  • FIG. 2 shows a schematic diagram of a nuclear reactor system in accordance with a first embodiment of the present invention
  • FIG. 3 shows a schematic diagram of a nuclear reactor system in accordance with a second embodiment of the present invention.
  • a nuclear reactor system 100 includes: a nuclear reactor 1 (an example of a heating device); a heat consuming device, a heat consuming device and a nuclear reactor 1 connected by a pipe to form a circuit 3; and steam,
  • the steam is in a wet steam state before being supplied to the nuclear reactor 1, and is supplied to the heat consuming device by heat exchange with the nuclear reactor 1 to become dry steam or superheated steam.
  • the heat consuming device may be the power generation system 7 shown in Fig. 2 or the heat exchanger 15 such as a steam generator shown in Fig. 3.
  • the steam can be formed from heavy water.
  • the nuclear reactor system 100 may further include: a steam-water separation device 6 disposed in the circuit 3 downstream of the steam outlet 101 of the nuclear reactor 1 for Liquid water, such as water droplets, in the steam output from the nuclear reactor 1 is separated.
  • a steam-water separation device 6 disposed in the circuit 3 downstream of the steam outlet 101 of the nuclear reactor 1 for Liquid water, such as water droplets, in the steam output from the nuclear reactor 1 is separated.
  • the nuclear reactor system 100 may further include: a humidity control device 12 disposed in the circuit 3 upstream of the steam inlet 102 of the nuclear reactor 1 for controlling the The humidity of the steam.
  • the nuclear reactor system 100 may further include: a temperature control device 11 disposed in the circuit 3 upstream of the humidity control device 12 for controlling the location The temperature of the steam.
  • the nuclear reactor system 100 may further include: a pressure control device 10 disposed in the circuit 3, upstream of the temperature control device 11, for controlling the The pressure of the steam.
  • the steam inlet 102 of the nuclear reactor 1 is disposed on the lower side of the nuclear reactor 1, and the steam outlet 101 of the nuclear reactor 1 is disposed on the upper side of the nuclear reactor 1, and the steam-water separation device 6 may It is disposed above the steam outlet 101.
  • the heat exchange system in nuclear reactor system 100 can convert heat into electrical energy in two ways.
  • the first mode is a direct power generation mode in which the steam in the circuit 3 directly drives the steam turbine to generate electricity as shown in FIG. 2, and the second mode is that the heat exchange between the circuit 3 and the circuit 4 is first performed as shown in FIG.
  • the steam in circuit 4 drives the indirect power generation of the turbine.
  • the nuclear reactor system 100 includes steam moisture
  • the steam-water separation device 6 may be located above the steam outlet 101 of the reactor 1 to separate the steam from the liquid droplets to prevent the liquid droplets from entering the steam turbine 7, causing damage to the blades of the steam turbine 7.
  • the boiling water reactor steam-water separation device may be selected, but the nuclear reactor system 100 may not Including the steam-water separation device 6, the nuclear reactor system can generate steam with high dryness; downstream of the steam-water separation device 6, a plurality of steam turbines 7 can be arranged as needed to form a steam turbine unit, and the steam turbine group can drive the generator set to generate electricity, which can be used at present Medium and low pressure steam turbines commonly used in reactor systems.
  • the steam replenishment system 8 is mainly used to supplement the steam in the circuit to ensure the normal operation of the circuit.
  • the function of the radioactive pollutant treatment system 9 is to treat pollutants such as impurities and radioactive vapors.
  • the pressure control device 10 functions to control the vapor pressure of the steam inlet 102 of the reactor 1, and the pressure control device 10 can be controlled by a high pressure boiler.
  • the temperature control device 11 is mainly used to adjust the steam temperature of the steam inlet 102 of the reactor 1, and the inside of the temperature control device 11 can adopt a tube bundle heating structure;
  • the function of the humidity control device 12 is to adjust the steam humidity of the steam inlet 102 of the reactor 1,
  • the function of the control valve 13 is to control the steam flow according to the pressure and temperature of the steam, and an induction high-pressure high-temperature control valve can be used.
  • the direct heat exchange system is mainly composed of the reactor 1 and the circuit 3. Its main function is to transfer the heat in the reactor 1 to the wet steam through heat exchange, and the wet steam is heat-exchanged to become dry steam or superheated steam. .
  • the steam heated by the heat enters the power generation system in the circuit 3, and directly drives the steam turbine 7 to generate electricity.
  • the cooling steam generated after power generation may first enter the radioactive pollutant treatment system 9 for the treatment of neutron toxicity fission products, and the disposed steam may sequentially enter the pressure control device 10, the temperature control device 11 and the humidity control device 12. The steam pressure is adjusted, the temperature is adjusted, and the humidity is adjusted.
  • the wet steam conforming to the standard enters the reactor 1 from the steam inlet 102 of the reactor 1, and enters the reactor inside the reactor 1 through the high temperature resistant high pressure pipeline.
  • the core channel transfers the heat inside the reactor 1 to the wet steam by heat exchange, and the wet steam is converted into dry steam or superheated steam by temperature rise, phase change, and is led out from the top of the reactor 1 and enters the loop 3 again. If it is detected that the water vapor in the heat exchange system is insufficient, the adjustable steam supply system 8 supplements the steam in the circuit to ensure the normal operation of the circuit.
  • the humidity of the wet steam in the circuit 3 may be 1% to 100%
  • the working pressure may be 1 MPa to 12 MPa
  • the working temperature may be 250 ° C to 950 ° C.
  • Steam formed by light water can be used as the heat exchange medium in the circuit 3.
  • the heat exchange system or nuclear reactor system 100 mainly includes a reactor 1, a circuit 3 and a circuit 4, specifically including a steam-water separation device 6, a steam turbine 7 for power generation, a steam replenishment system 8, and radioactive pollution.
  • the steam generator 15 is a device for realizing the heat transfer between the circuit 3 and the circuit 4, which uses the dry steam of the heavy water in the circuit 3 or the superheated steam to heat the steam in the circuit 4, so that the steam in the circuit 4 becomes high-temperature steam, and the high-temperature steam is The circuit 4 pushes the steam turbine 7 to generate electricity.
  • the heat-exchanged cooling steam in the circuit 3 first enters the radioactive pollutant treatment system 9 for the treatment of neutron toxicity fission products, and the disposed steam sequentially enters the pressure control device 10, the temperature control device 11, and the humidity control device 12
  • the wet steam conforming to the standard after reforming of various parameters enters the reactor 1 from the steam inlet 102 below the reactor 1, and enters the reactor through the high temperature resistant high pressure pipeline.
  • the core channel transfers the heat in the reactor to the wet steam by heat exchange, and the wet steam is converted into dry steam or superheated steam by temperature rise, phase change, and is led out from the top of the reactor 1 and enters the steam generator 15 again.
  • the adjustable steam supply system 8 supplements the steam in the circuit 4 to ensure the normal operation of the circuit 4.
  • the power generation system 7 is mainly composed of a plurality of medium and low pressure steam turbines 7, and its main function is to generate electricity by using high temperature steam under different pressures.
  • steam formed by heavy water can be used as the heat exchange medium
  • steam formed by light water can be used as the heat exchange medium.
  • the nuclear reactor system can operate as a critical reactor system (as shown in the solid lines in Figures 2 and 3) or as a subcritical core or cladding driven by an external neutron source. Run (as shown in the solid and dashed lines in Figures 2 and 3).
  • the wet steam as the reactor coolant is heated to about 200 degrees Celsius in the temperature control device 11, and the pressure in the pressure control device 10 is adjusted to 3-12 MP, and at 10-70 m/s.
  • a spalling target 19 that generates a driving neutron source is passed through the reactor pressure vessel 20 and the core 21, which is inside the core 21 and has a closed structure to prevent spalling.
  • the spalling medium in the target is in contact with the core coolant.
  • the heat deposition generated by the spallation target 19 and the beam coupling is exchanged by the spalling target heat exchange system 14, and the loop 5 is independent of the loop 3 of the nuclear reactor system 100.
  • no external drive is required, and the fission is self-sustained.
  • the nuclear reactor 1 has a fuel rod 16.
  • the reactor heat exchange medium wet steam enters the passage 17 in the reactor through the bottom inlet 102 of the reactor 1, exchanges heat therewith, removes heat, and cools the reactor.
  • the principle of the fuel cycle system 200 is illustrated in Figure 5, which contains both processing and combustion of spent fuel (core material after reactor combustion).
  • the spent fuel (or depleted uranium, natural uranium, thorium, etc.) produced in the nuclear reactor system is treated by a simple high-temperature dry process to remove neutron-toxic nuclides from spent fuel, and the remaining spent material is made into components and placed in a burner. (reactor) burning.
  • the burner expands the fuel while reducing the MA content to form a new fuel, so many cycles. No excess radioactive waste is produced in the flow of the fuel cycle system, and capacity can be produced while the nuclear waste is metamorphosed and proliferated.
  • the fuel cycle system 200 in accordance with an embodiment of the present invention excludes only about 50% of the fission products during spent fuel treatment, and the lanthanides remain in the fuel to continue combustion.
  • the separation difficulty and cost are greatly reduced.
  • the total amount of nuclear waste discharged is also greatly reduced ( ⁇ 4% of total spent fuel), and the radiotoxicity is greatly reduced (the content of MA is less than 0.1% of the original content of spent fuel).
  • the internal heat exchange pressure of the reactor in the heat exchange system can be lower than that of the pure gas cooling system; and the relative purity of the pure water system, the wet steam as the cooling medium, the safety of the system and The controllability is higher; the fission reactor using the heat exchange medium of the embodiment of the invention is suitable for the fast neutron or ultrafast neutron spectrum, can meet the requirements of high power density, and can use uranium 235, strontium, uranium 238 Long-lived fission products, transuranic elements as nuclear fuel, and can be used for the metamorphosis of nuclear waste and isotope production.
  • the wet steam according to the embodiment of the present invention is used as a cooling medium. Compared with the original single-phase medium, since the wet steam itself can undergo a phase change due to heat, the heat exchange effect can be better improved, and at the same time, the partial pressure can also be achieved. Adjustment to control heat transfer efficiency.
  • a Brayton cycle can be combined with a Rankine cycle.
  • the circuit 4 can adopt a standard water circuit, similar to the current pressurized water reactor circuit.
  • the steam is a medium in which a gas and a liquid are simultaneously present in a specific space.
  • the gaseous vapor also includes superheated steam.
  • the vapor may have a density of from 1 g/m 3 to 80 g/m 3 .
  • the core material may be a SiC composite.
  • the water vapor cooling medium has the advantages of large heat capacity, low pressure system, non-corrosive, off-line processing, and the like. This kind of water vapor cooling medium cooled fission reactor can operate safely and reliably at high power density.
  • the heat exchange system according to the invention can also be used for heat exchange between other heating devices and heat consumers.

Abstract

一种热交换系统和核反应堆系统包括：加热装置（1）；热量消耗装置，热量消耗装置与加热装置（1）通过管道连接而形成回路（3）；以及蒸汽，所述蒸汽在供给加热装置（1）前处于湿蒸汽状态，并且通过与加热装置（1）热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。上述系统通过采用水蒸汽作为换热介质提高了换热效率，并改善了核反应堆系统的安全性。

Description

热交换系统和核反应堆系统 技术领域
本发明涉及一种热交换系统和核反应堆系统。
背景技术
核反应堆系统通常采用液态金属作为冷却介质。
发明内容
本发明的目的是提供一种热交换系统和核反应堆系统，通过采用水蒸汽作为换热介质提高了换热效率，并改善了核反应堆系统的安全性。
根据本发明的实施例，提供了一种热交换系统，包括：加热装置；热量消耗装置，热量消耗装置与加热装置通过管道连接而形成回路；以及蒸汽，所述蒸汽在供给热源前处于湿蒸汽状态，并且通过与加热装置热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。
根据本发明的实施例，所述的热交换系统还包括：汽水分离装置，所述汽水分离装置设置在所述回路中、所述加热装置的蒸汽出口的下游，用于分离出从加热装置输出的蒸汽中的液态水。
根据本发明的实施例，所述的热交换系统还包括：湿度控制装置，所述湿度控制装置设置在所述回路中、加热装置的蒸汽入口的上游，用于控制所述蒸汽的湿度。
根据本发明的实施例，所述的热交换系统还包括：温度控制装置，所述温度控制装置设置在所述回路中、所述湿度控制装置的上游，用于控制所述蒸汽的温度。
根据本发明的实施例，所述的热交换系统还包括：压力控制装置，所述压力控制装置设置在所述回路中、所述温度控制装置的上游，用于控制所述蒸汽的压力。
根据本发明的实施例，所述热量消耗装置是换热器或发电系统。
根据本发明的实施例，所述蒸汽由重水形成。
根据本发明的实施例，所述加热装置的蒸汽入口设置在所述加热装置的下侧，并且所述加热装置的蒸汽出口设置在所述加热装置的上侧，所述汽水分离装置设置在所述蒸汽出口的上方。
根据本发明的实施例，提供了一种核反应堆系统，该核反应堆系统包括：核反应堆；热量消耗装置，热量消耗装置与核反应堆通过管道连接而形成回路；以及蒸汽，所述蒸汽在供给核反应堆前处于湿蒸汽状态，并且通过与核反应堆热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。
根据本发明的实施例，所述的核反应堆系统还包括：汽水分离装置，所述汽水分离装置设置在所述回路中、所述核反应堆的蒸汽出口的下游，用于分离出从核反应堆输出的蒸汽中的液态水。
在根据本发明的实施例中，水蒸汽作为冷却介质具有大热容量、低压系统、无腐蚀、离线处理等优势。此种水蒸汽冷却介质冷却的裂变反应堆可以在高功率密度下安全可靠运行。
附图说明
图1为根据本发明的实施例的核能系统的示意图；
图2为根据本发明的第一实施例的核反应堆系统的示意图；
图3为根据本发明的第二实施例的核反应堆系统的示意图；
图4为根据本发明的实施例的反应堆的示意图；以及
图5为根据本发明的实施例的产能的燃料循环示意图。
具体实施方式
下面结合附图及具体实施方式对本发明做进一步说明。
图1示出了根据本发明的实施例的核能系统的示意图。如图1所示，根据本发明的示例性实施例的核能系统包括：核反应堆系统100和燃料循环系统200。核能系统可以是快中子核能系统。核反应堆系统100可以是快中子核反应堆系统。
图2示出了根据本发明的第一实施例的核反应堆系统的示意图，而图3示出了根据本发明的第二实施例的核反应堆系统的示意图。
如图2和3所示，根据本发明的实施例的核反应堆系统100包括：核反应堆1(加热装置的示例)；热量消耗装置，热量消耗装置与核反应堆1通过管道连接而形成回路3；以及蒸汽，所述蒸汽在供给核反应堆1前处于湿蒸汽状态，并且通过与核反应堆1热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。所述热量消耗装置可以是图2所示的发电系统7或图3所示的诸如蒸汽发生器的换热器15。所述蒸汽可以由重水形成。
如图2和3所示，所述的核反应堆系统100还可以包括：汽水分离装置6，所述汽水分离装置6设置在所述回路3中、所述核反应堆1的蒸汽出口101的下游，用于分离出从核反应堆1输出的蒸汽中的液态水，例如水滴。
如图2和3所示，所述的核反应堆系统100还可以包括：湿度控制装置12，所述湿度控制装置12设置在所述回路3中、核反应堆1的蒸汽入口102的上游，用于控制所述蒸汽的湿度。
如图2和3所示，所述的核反应堆系统100还可以包括：温度控制装置11，所述温度控制装置11设置在所述回路3中、所述湿度控制装置12的上游，用于控制所述蒸汽的温度。
如图2和3所示，所述的核反应堆系统100还可以包括：压力控制装置10，所述压力控制装置10设置在所述回路中3、所述温度控制装置11的上游，用于控制所述蒸汽的压力。
如图2和3所示，所述核反应堆1的蒸汽入口102设置在所述核反应堆1下侧，并且所述核反应堆1的蒸汽出口101设置在所述核反应堆1上侧，所述汽水分离装置6可以设置在所述蒸汽出口101的上方。
如图2和3所示，核反应堆系统100中的热交换系统可以采用两种方式将热量转换成电能。第一种方式为如图2所示的由回路3中的蒸汽直接推动汽轮机发电的直接发电方式，第二种方式为如图3所示的先通过回路3与回路4进行热交换，再利用回路4中的蒸汽推动汽轮机发电的间接发电方式。
在图2所示的实施例中，所述的核反应堆系统100包括汽水分 离装置6、用于发电的汽轮机7、水蒸汽补给系统8，放射性污染物处理系统9、压力控制装置10、湿度控制装置11、温度控制装置12、控制阀13以及耐高温、高压的管道。汽水分离装置6可以位于反应堆1蒸汽出口101上方，作用是把蒸汽和液滴分开，防止液滴进入汽轮机7，造成汽轮机7叶片损坏，其可选用沸水堆汽水分离装置，但核反应堆系统100可以不包括汽水分离装置6，核反应堆系统可以产生干度很高的蒸汽；在汽水分离装置6的下游根据需要可设置有多个汽轮机7，形成汽轮机组，汽轮机组可驱动发电机组发电，其可采用目前反应堆系统中常用的中、低压汽轮机组。水蒸汽补给系统8主要用于补充回路中的蒸汽，保证回路的正常运行。放射性污染物处理系统9的作用是处理带有杂质、放射性的蒸汽等污染物，压力控制装置10的作用是控制反应堆1的蒸汽入口102的蒸汽压力，压力控制装置10可采用高压锅炉进行压力控制；温度控制装置11主要用于调节反应堆1的蒸汽入口102的蒸汽温度，温度控制装置11的内部可采用管束加热结构；湿度控制装置12的作用是调节反应堆1的蒸汽入口102的蒸汽湿度，可采用喷雾式控制方法，控制阀13的作用是根据蒸汽的压力、温度控制蒸汽流量，可采用感应式高压高温控制阀。
如图2所示，直接热交换系统主要由反应堆1和回路3组成，其主要功能是将反应堆1内的热量通过热交换传递到湿蒸汽，湿蒸汽进行热交换后成为干蒸汽或者为过热蒸汽。受热升温后的蒸汽进入回路3中的发电系统，直接推动汽轮机7发电。发电后所产生的冷却蒸汽首先可进入放射性污染物处理系统9中进行中子毒性裂变产物的处置，处置完后的蒸汽可依次进入到压力控制装置10、温度控制装置11和湿度控制装置12中，进行蒸汽压力的调整，温度的调节以及湿度的调控，经过各种参数的重整后符合标准的湿蒸汽从反应堆1的蒸汽入口102进入反应堆1，通过耐高温高压管道进入反应堆1内部的堆芯通道，将反应堆1内部热量通过热交换的方式转移到湿蒸汽，湿蒸汽经过升温、相变转换为干蒸汽或者过热蒸汽，从反应堆1顶部导出，再次进入回路3。若检测出该换热系统中的水蒸汽不足，可调用水蒸汽补给系统8补充回路中蒸汽，保证回路的正常运行。根据本 发明的实施例，回路3中的湿蒸汽的湿度可为1％～100％，工作压强可为1MPa～12MPa，工作温度可为250℃～950℃。回路3中可以采用轻水形成的蒸汽作为换热介质。
在图3所示的实施例中，热交换系统或核反应堆系统100主要包括反应堆1、回路3和回路4，具体包括汽水分离装置6、用于发电的汽轮机7、水蒸汽补给系统8，放射性污染物处理系统9、压力控制装置10、温度控制装置11、湿度控制装置12、控制阀13、用于回路3和回路4热交换的蒸汽发生器15以及耐高温、高压的管道等。蒸汽发生器15是实现回路3与回路4热量传递的设备，它利用回路3中的重水的干蒸汽或者过热蒸汽加热回路4中的蒸汽，使回路4中的蒸汽变为高温蒸汽，高温蒸汽在回路4中推动汽轮机7发电。回路3中的换热后的冷却蒸汽首先进入放射性污染物处理系统9中进行中子毒性裂变产物的处置，处置完后的蒸汽依次进入到压力控制装置10、温度控制装置11和湿度控制装置12中，进行蒸汽压力的调整，温度的调节以及湿度的调控，经过各种参数的重整后符合标准的湿蒸汽从反应堆1下方的蒸汽入口102进入反应堆1，通过耐高温高压管道进入反应堆内部的堆芯通道，将反应堆内热量通过热交换的方式转移到湿蒸汽，湿蒸汽经过升温、相变转换为干蒸汽或者过热蒸汽，从反应堆1顶部导出，再次进入蒸汽发生器15。若检测出该换热系统中的水蒸汽不足，可调用水蒸汽补给系统8补充回路4中的蒸汽，保证回路4的正常运行。其中发电系统7主要由多个中、低压汽轮机7组成，其主要功能是利用不同压强下的高温蒸汽进行发电。回路3中可以采用重水形成的蒸汽作为换热介质，而回路4中可以采用轻水形成的蒸汽作为换热介质。
如图2和图3所示，该核反应堆系统可以作为临界反应堆系统运行(如图2和3中实线部分所示)，也可以作为外中子源驱动下的次临界堆芯或包层来运行(如图2和3中实线和虚线部分所示)。
在作为临界反应堆运行过程中，作为反应堆冷却剂的湿蒸汽在温度控制装置11中被加热到200摄氏度左右，在压力控制装置10内压力被调节到3-12MP，并以10-70m/s的速度进入堆芯，在堆芯通过 热交换，出口温度能够达到400-950摄氏度。此种换热方式能够保证堆芯可以较高功率密度运行。
如图4所示，在作为次临界反应堆系统运行过程中，产生驱动中子源的散裂靶19贯穿反应堆压力容器20和堆芯21，其在堆芯21内部且为密闭结构，杜绝散裂靶中散裂工质与堆芯冷却剂相接触。散裂靶19和束流耦合产生的热量沉积通过散裂靶换热系统14进行换热，其回路5与核反应堆系统100的回路3相互独立。在临界反应堆运行过程中，不需要外源驱动装置，即可自持裂变。核反应堆1具有燃料棒16。反应堆换热介质湿蒸汽通过反应堆1底部入口102进入反应堆中的通道17，与其进行热交换，带走热量，进行反应堆的冷却。
燃料循环系统200的原理如图5所示，该系统包含乏燃料(经过反应堆燃烧过后的核材料)的处理和燃烧两大部分。核反应堆系统中产出的乏燃料(或者贫铀、天然铀、钍等)通过简单的高温干法处理，排除乏燃料中的中子毒性核素，将剩余的乏料制成元件并放入燃烧器(反应堆)燃烧。燃烧器在减小MA含量的同时进行燃料的增殖形成新燃料，如此多次循环。在燃料循环系统的流程中不产生多余的放射废料，并可在核废料嬗变和增殖的同时产能。
根据本发明的实施例的燃料循环系统200在乏燃料处理过程中，仅排除掉约50％的裂变产物，锕系元素仍保留在燃料中继续燃烧。大大降低了分离难度和费用。同时，排放的核废料总量也大为减少(＜乏燃料总量的4％)，放射性毒性大为降低(MA的含量小于乏燃料中原含量的0.1％)。
采用湿蒸汽作为冷却介质的情况下，热交换系统中反应堆的内部热交换压强可以比纯气冷却系统具有更低的压强；且相对纯水系统，湿蒸汽作为冷却工质，系统的安全性和可控性更高；使用本发明的实施例的热交换介质的裂变反应堆适用于快中子或超快中子谱的场合，可以满足高功率密度的要求，可以使用铀235、钍、铀238、长寿命裂变产物、超铀元素作为核燃料，并可以用于核乏料的嬗变和同位素生产。
采用根据本发明的实施例的湿蒸汽作为冷却介质，相对于原有单相介质，由于湿蒸汽本身可以受热发生相变，故能更好的提高换热效果，同时，也可通过对局部压强的调节来控制换热效率。
对于核反应堆系统的直接发电方式，可以采取布雷顿循环与朗肯循环结合。
对于核反应堆系统的如图3所示的间接发电方式，回路4可以采取标准的水回路，类似目前压水堆的回路。
所述蒸汽，在特定空间以气、液二态同时存在的一种介质。所述气态蒸汽也包括过热蒸汽。所述蒸汽的密度可以为1g/m3～80g/m3。所述堆芯材料可以为SiC复合材料。
在根据本发明的实施例中，水蒸汽冷却介质具有大热容量、低压系统、无腐蚀、离线处理等优势。此种水蒸汽冷却介质冷却的裂变反应堆可以在高功率密度下安全可靠运行。
此外，根据本发明的热交换系统也可以用于其它的加热装置和热量消耗装置之间的换热。

Claims (10)

  1. 一种热交换系统，包括：
    加热装置；
    热量消耗装置，热量消耗装置与加热装置通过管道连接而形成回路；以及
    蒸汽，所述蒸汽在供给热源前处于湿蒸汽状态，并且通过与加热装置热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。
  2. 根据权利要求1所述的热交换系统，还包括：
    汽水分离装置，所述汽水分离装置设置在所述回路中、所述加热装置的蒸汽出口的下游，用于分离出从加热装置输出的蒸汽中的液态水。
  3. 根据权利要求1所述的热交换系统，还包括：
    湿度控制装置，所述湿度控制装置设置在所述回路中、加热装置的蒸汽入口的上游，用于控制所述蒸汽的湿度。
  4. 根据权利要求3所述的热交换系统，还包括：
    温度控制装置，所述温度控制装置设置在所述回路中、所述湿度控制装置的上游，用于控制所述蒸汽的温度。
  5. 根据权利要求4所述的热交换系统，还包括：
    压力控制装置，所述压力控制装置设置在所述回路中、所述温度控制装置的上游，用于控制所述蒸汽的压力。
  6. 根据权利要求1所述的热交换系统，其中：
    所述热量消耗装置是换热器或发电系统。
  7. 根据权利要求1所述的热交换系统，其中：
    所述蒸汽由重水形成。
  8. 根据权利要求2所述的热交换系统，其中：
    所述加热装置的蒸汽入口设置在所述加热装置的下侧，并且所述加热装置的蒸汽出口设置在所述加热装置的上侧，所述汽水分离装置设置在所述蒸汽出口的上方。
  9. 一种核反应堆系统，包括：
    核反应堆；
    热量消耗装置，热量消耗装置与核反应堆通过管道连接而形成回路；以及
    蒸汽，所述蒸汽在供给核反应堆前处于湿蒸汽状态，并且通过与核反应堆热交换成为干蒸汽或者过热蒸汽后供给热量消耗装置。
  10. 根据权利要求9所述的核反应堆系统，还包括：
    汽水分离装置，所述汽水分离装置设置在所述回路中、所述核反应堆的蒸汽出口的下游，用于分离出从核反应堆输出的蒸汽中的液态水。
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Families Citing this family (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US9433318B2 (en) 2013-08-30 2016-09-06 Seattle Espresso Machine Corporation System and apparatus for regulating flow rate in an espresso machine
US9364117B2 (en) 2013-08-30 2016-06-14 Seattle Espresso Machine Corporation Method for regulating flow rate in an espresso machine
ES2873507T3 (es) * 2016-05-19 2021-11-03 European Spallation Source Eric Método para proporcionar una fuente de neutrones
US9907426B2 (en) * 2016-05-20 2018-03-06 Seattle Espresso Machine Corporation Method for generation of superheated steam for the preparation of a beverage
US10390654B2 (en) 2016-05-20 2019-08-27 Seattle Espresso Machine Corporation Generation of superheated steam for the preparation of a beverage
CN107146641A (zh) * 2017-05-11 2017-09-08 中国科学院近代物理研究所 核能系统和控制核能系统的方法
CN111779576B (zh) * 2020-07-13 2022-07-05 中国航空发动机研究院 一种组合式推进装置、系统及控制方法
GB202207761D0 (en) * 2022-05-26 2022-07-13 Egb Eng Consultants Ltd A nuclear reactor cycle

Citations (10)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB954529A (en) * 1961-12-11 1964-04-08 Atomic Energy Commission Steam-water mixing device for steam-cooled reactor
GB1174796A (en) * 1966-03-31 1969-12-17 Reactor Centrum Nederland A Steam-Cooled Nuclear Reactor, Especially of the Epithermal or of the Fast Type
US4714055A (en) * 1984-09-17 1987-12-22 Framtome & Cie Water and steam separating device for drying moist steam
CN1914463A (zh) * 2003-12-16 2007-02-14 先进燃烧能量系统公司 生产能量的设备和方法
CN101299351A (zh) * 2008-06-27 2008-11-05 张育曼 水冷双区增殖核反应堆堆芯及采用该堆芯的核反应堆
CN101719385A (zh) * 2009-12-08 2010-06-02 华北电力大学 超导热管式核电热冷联产系统
CN101908385A (zh) * 2010-07-02 2010-12-08 华北电力大学 利用盐溶液吸湿特性缓解核电站严重事故的装置
CN103208317A (zh) * 2012-01-17 2013-07-17 巴布科克和威尔科克斯核能股份有限公司 具有外部蒸汽鼓筒的一体式压水反应堆
CN103544997A (zh) * 2013-10-28 2014-01-29 中科华核电技术研究院有限公司 定位格架及反应堆燃料组件
CN104081466A (zh) * 2011-12-12 2014-10-01 控制工程学公司 微粒去除系统

Family Cites Families (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
NL236115A (zh) * 1958-02-14 1900-01-01
CH382311A (de) * 1960-02-27 1964-09-30 Sulzer Ag Verfahren zum Betrieb einer Wärmekraftanlage mit einem Atomkernreaktor
GB1116479A (en) * 1966-12-27 1968-06-06 Exxon Research Engineering Co Removal of heat from nuclear reactors
US3400048A (en) * 1967-01-11 1968-09-03 Gen Electric Steam cooled nuclear reactor power system with steam decontamination trament
US3400049A (en) * 1967-01-11 1968-09-03 Gen Electric Steam cooled nuclear reactor power system
CN101807443A (zh) * 2010-03-25 2010-08-18 华北电力大学 一种基于压水堆和高温气冷堆的混合热力循环系统
CN103760108B (zh) * 2013-12-30 2017-01-25 浙江大学 一种湿蒸汽湿度的光声差动测量方法与装置

Patent Citations (10)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
GB954529A (en) * 1961-12-11 1964-04-08 Atomic Energy Commission Steam-water mixing device for steam-cooled reactor
GB1174796A (en) * 1966-03-31 1969-12-17 Reactor Centrum Nederland A Steam-Cooled Nuclear Reactor, Especially of the Epithermal or of the Fast Type
US4714055A (en) * 1984-09-17 1987-12-22 Framtome & Cie Water and steam separating device for drying moist steam
CN1914463A (zh) * 2003-12-16 2007-02-14 先进燃烧能量系统公司 生产能量的设备和方法
CN101299351A (zh) * 2008-06-27 2008-11-05 张育曼 水冷双区增殖核反应堆堆芯及采用该堆芯的核反应堆
CN101719385A (zh) * 2009-12-08 2010-06-02 华北电力大学 超导热管式核电热冷联产系统
CN101908385A (zh) * 2010-07-02 2010-12-08 华北电力大学 利用盐溶液吸湿特性缓解核电站严重事故的装置
CN104081466A (zh) * 2011-12-12 2014-10-01 控制工程学公司 微粒去除系统
CN103208317A (zh) * 2012-01-17 2013-07-17 巴布科克和威尔科克斯核能股份有限公司 具有外部蒸汽鼓筒的一体式压水反应堆
CN103544997A (zh) * 2013-10-28 2014-01-29 中科华核电技术研究院有限公司 定位格架及反应堆燃料组件

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