WO2011075923A1 - 高温气冷堆蒸汽发电系统及方法 - Google Patents

高温气冷堆蒸汽发电系统及方法 Download PDF

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Publication number
WO2011075923A1
WO2011075923A1 PCT/CN2010/000085 CN2010000085W WO2011075923A1 WO 2011075923 A1 WO2011075923 A1 WO 2011075923A1 CN 2010000085 W CN2010000085 W CN 2010000085W WO 2011075923 A1 WO2011075923 A1 WO 2011075923A1
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WO
WIPO (PCT)
Prior art keywords
steam
pressure cylinder
temperature gas
power generation
cooled reactor
Prior art date
Application number
PCT/CN2010/000085
Other languages
English (en)
French (fr)
Inventor
张作义
吴宗鑫
王大中
徐元辉
孙玉良
李富
董玉杰
Original Assignee
清华大学
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 清华大学 filed Critical 清华大学
Priority to KR1020127016026A priority Critical patent/KR101454089B1/ko
Priority to EP10838490.0A priority patent/EP2518733B1/en
Priority to PL10838490T priority patent/PL2518733T3/pl
Priority to US13/519,109 priority patent/US9111652B2/en
Priority to JP2012545049A priority patent/JP5645283B2/ja
Priority to BR112012015552-0A priority patent/BR112012015552B1/pt
Priority to CA2785255A priority patent/CA2785255C/en
Priority to RU2012126055/07A priority patent/RU2515496C2/ru
Publication of WO2011075923A1 publication Critical patent/WO2011075923A1/zh
Priority to ZA2012/04897A priority patent/ZA201204897B/en

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Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D5/00—Arrangements of reactor and engine in which reactor-produced heat is converted into mechanical energy
    • G21D5/04—Reactor and engine not structurally combined
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F01—MACHINES OR ENGINES IN GENERAL; ENGINE PLANTS IN GENERAL; STEAM ENGINES
    • F01K—STEAM ENGINE PLANTS; STEAM ACCUMULATORS; ENGINE PLANTS NOT OTHERWISE PROVIDED FOR; ENGINES USING SPECIAL WORKING FLUIDS OR CYCLES
    • F01K7/00—Steam engine plants characterised by the use of specific types of engine; Plants or engines characterised by their use of special steam systems, cycles or processes; Control means specially adapted for such systems, cycles or processes; Use of withdrawn or exhaust steam for feed-water heating
    • F01K7/16—Steam engine plants characterised by the use of specific types of engine; Plants or engines characterised by their use of special steam systems, cycles or processes; Control means specially adapted for such systems, cycles or processes; Use of withdrawn or exhaust steam for feed-water heating the engines being only of turbine type
    • F01K7/22—Steam engine plants characterised by the use of specific types of engine; Plants or engines characterised by their use of special steam systems, cycles or processes; Control means specially adapted for such systems, cycles or processes; Use of withdrawn or exhaust steam for feed-water heating the engines being only of turbine type the turbines having inter-stage steam heating
    • 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
    • F22B1/18—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being a hot gas, e.g. waste gas such as exhaust gas of internal-combustion engines
    • F22B1/1823—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being a hot gas, e.g. waste gas such as exhaust gas of internal-combustion engines for gas-cooled nuclear reactors
    • 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
    • G21C1/07—Pebble-bed reactors; Reactors with granular fuel
    • 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
    • G21C1/10—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 moderator and coolant being different or separated
    • G21C1/12—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 moderator and coolant being different or separated moderator being solid, e.g. Magnox reactor or gas-graphite reactor
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C5/00—Moderator or core structure; Selection of materials for use as moderator
    • G21C5/02—Details
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D5/00—Arrangements of reactor and engine in which reactor-produced heat is converted into mechanical energy
    • G21D5/04—Reactor and engine not structurally combined
    • G21D5/08—Reactor and engine not structurally combined with engine working medium heated in a heat exchanger by the reactor coolant
    • G21D5/12—Liquid working medium vaporised by reactor coolant
    • 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 the field of nuclear power, and in particular to a high temperature gas cooled reactor steam power generation system and method. Background technique
  • nuclear power generation is important for mitigating energy security and global climate change.
  • US San Francisco and the former Soviet Chernobyl nuclear power plant people are still actively developing safer and more economical nuclear power generation technologies.
  • the third generation of nuclear power technology is basically mature.
  • high-temperature gas-cooled reactors can achieve high export temperatures, high power generation efficiency, and high-grade heat energy supply, which have caused widespread concern.
  • the high-temperature gas-cooled reactor uses a ceramic-coated granular fuel element, helium gas is used as a coolant, and graphite is used as a moderator.
  • the core outlet temperature can reach 700 ° C to 950 ° C.
  • the high temperature gas cooled reactor is a safe pile with the following reasons: 1) excellent fuel element performance; 2) large heat capacity of the graphite core; 3) full range of negative reactivity temperature coefficient; 4) coolant ⁇
  • the gas is an inert gas with good chemical stability and no phase change.
  • the modular high-temperature gas-cooled reactor refers to a high-temperature gas-cooled reactor with inherent safety characteristics and a small single-stage power.
  • the basic characteristics of this reactor are: Reaction under any accident conditions The residual heat of the core can be carried out in a passive manner, and the maximum temperature of the core fuel does not exceed the allowable limit. Since the possibility of core melting is avoided, even if a very low probability of overdesign basis accident occurs, the radioactive dose outside the nuclear power plant is still within the limits, and it is technically unnecessary to draw an off-site contingency plan.
  • the high temperature gas cooled reactor is divided into a ball bed pile and a column pile.
  • the former compresses the coated particulate fuel together with the graphite matrix into a fuel ball of 6 cm in diameter to form a flowable core of the ball bed, thereby realizing the online replacement of nuclear fuel without stopping the pile.
  • the latter compresses the coated particulate fuel together with graphite into a cylindrical pellet which is then placed in a hexagonal prismatic fuel assembly to form a fixed prismatic core.
  • the ball-bed type high-temperature gas-cooled reactor has the following characteristics: 1) The fuel element is not loaded and unloaded, and the power station has high availability; 2) the excess reactivity of the core is small, the reaction control is convenient, and the neutron economy is high; 3) Uniform fuel consumption, high fuel discharge and high fuel utilization; 4) Low fuel pellet temperature during normal operation, which is easy to further increase the reactor outlet temperature.
  • modular high temperature gas cooled reactors are primarily due to safety considerations.
  • the inherent safety requirements of modular high-temperature gas-cooled reactors require that the post-acceleration decay heat can be used to derive the core in a passive manner, ensuring that the maximum fuel temperature does not exceed the design limit, technically proposed for the power density and total power of a single core. limit.
  • An object of the present invention is to provide a high temperature gas cooled reactor steam power generation system and method which can overcome the drawbacks of the prior art and achieve economy while ensuring safety.
  • a high temperature gas cooled reactor steam power generation system comprising: a plurality of nuclear steam supply systems, high pressure cylinders, low pressure cylinders, condensers that are sequentially connected end to end to form a closed steam loop , condensate pumps, low pressure heaters, deaerators, feed pumps and high pressure heaters.
  • the high temperature gas cooled reactor steam power generation system sequentially connects the steam reheater and the intermediate pressure cylinder between the high pressure cylinder and the low pressure cylinder.
  • the outlet of the high pressure heater is connected to a preliminary heating section of the steam reheater, and the inlet of the steam generator is connected to a preliminary heating section of the steam reheater.
  • the outlet of the high pressure cylinder is connected to the reheating section of the steam generator.
  • the high pressure cylinders are respectively connected to a reheater and an intermediate pressure cylinder, the outlet of the intermediate pressure cylinder is connected to a reheater, and the reheater is connected to the low pressure cylinder.
  • the nuclear steam supply system comprises a reactor and a steam generator respectively disposed in two pressure vessels, the reactor and the steam generator being connected by a hot gas conduit, and the upper part of the steam generator is arranged There is a main fan.
  • the reactor has a core designed as a flowable pebble bed structure with fuel elements located within the core and flowing from the top of the core to the bottom of the core.
  • the reactor has a core of a fixedly arranged prismatic structure with fuel elements located within the core.
  • the fuel element is coated with an all-ceramic type coated particulate fuel element.
  • the steam generator is a direct current steam generator, and the spiral tube structure is used.
  • the hot gas conduit is configured by a ring structure, the outer ring is a cold helium flow path for flowing helium from the steam generator to the reactor, and the inner ring is a heat for flowing helium from the reactor to the steam generator. Airflow path.
  • the invention also provides a high temperature gas cooled reactor steam power generation method, which comprises the steps of:
  • the steam is connected in parallel and then sent to a high pressure cylinder and a low pressure cylinder to perform work to drive power generation.
  • step S2 after the steam is sent to the high pressure cylinder for work, the steam flowing out of the high pressure cylinder enters the steam reheater for heating, and then enters the intermediate pressure cylinder and the low pressure cylinder work.
  • step S3 the wet steam that has been completed is heated to enter the preliminary heating section of the steam reheater before entering the steam generator.
  • step S2 after the steam is sent to the high pressure cylinder for work, the steam flowing out of the high pressure cylinder enters the reheating portion of the steam generator for heating, and then enters the intermediate pressure cylinder and the low cylinder to perform work.
  • step S2 after the steam is sent to the high-pressure cylinder for work, a part of the steam flowing out from the high-pressure cylinder enters the reheater and directly heats, and another part of the steam flowing out from the high-pressure cylinder first enters the intermediate-pressure cylinder to perform work. Then, the outlet steam is heated by the reheater, and finally the steam which is directly heated and firstly worked by the intermediate pressure cylinder and then heated is sent to the low pressure cylinder for work.
  • a reactor core and a pressure vessel are equipped with a steam generator as a standard module to form a nuclear steam supply system (NSSS) module.
  • the NSSS modules are replicated to provide steam for a large steam turbine power generation system, ie multiple NSSS modules with a steam turbine to achieve a "multi-NSSS module with one machine” configuration mode. Smaller individual modules reduce manufacturing complexity, and NSSS modules reduce cost due to bulk copying.
  • the NSSS module shares some auxiliary systems to improve the utilization of the auxiliary system and further reduce costs.
  • Multiple “multi-NSSS modules with one machine” generator set can be configured in one site to further share power plant auxiliary facilities and reduce construction and operation costs.
  • NSS nuclear steam supply system
  • NSS nuclear steam supply system
  • Figure 3 is a schematic structural view of an embodiment of a high temperature gas cooled reactor steam power generation system of the present invention
  • Figure 4 is a schematic view showing the structure of another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • Figure 5 is a schematic view showing the structure of still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • Figure 6 is a schematic view showing the structure of still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • Fig. 7 is a schematic view showing the structure of still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • 1 reactor; 2: core; 3: cold air flow; 4: hot air flow; 5: core top; 6: core bottom; 7: high temperature and high pressure steam; 8: second circuit water; : steam generator; 10: main blower; 11: NSSS module; 12: nuclear power plant auxiliary system; 13: steam power system; 14: generator; 15: steam reheater; 21: high pressure cylinder; 21: low pressure cylinder; 23: Condenser; 24: Condensate pump; 25: Low pressure heater; 26: Deaerator; 27: Feed water pump; 28: High pressure heater; 29: Medium pressure cylinder; 30: Reheater; 33: Fuel element.
  • FIG. 1 A steam power generation system having a nuclear steam supply system (NSSS) module of the present invention is shown in FIG.
  • a reactor core and a pressure vessel are equipped with a steam generator as a standard module to form a nuclear steam supply system (NSSS) module.
  • the nuclear power plant auxiliary system 12 mainly includes: a fuel loading and unloading system, a primary circuit pressure relief system, a helium purification and helium assist system, a gas sampling and analysis system, a residual heat exhaust system, a steam generator accident discharge system, a device cooling water system, Reactor plant ventilation and air conditioning systems, liquid waste treatment systems, solid waste treatment and storage systems, nuclear island fire protection systems, etc.
  • FIG. 2 is a schematic diagram showing the structure of a nuclear steam supply system (NSSS) module in accordance with an embodiment of the present invention.
  • the reactor 1 and the steam generator 9 are respectively disposed in two pressure vessels, which are connected by a hot gas conduit 32 to form a "side by side” arrangement.
  • the pressure vessel of the reactor 1, the casing of the steam generator 9, and the casing of the hot gas conduit 32 form a circuit pressure boundary and are installed in a concrete shielded compartment.
  • the hot gas conduit 32 has a ring-shaped structure, and the inner ring is a hot gas flow path 4, and the flow direction is from the reactor 1 to the steam generator 9.
  • the outer ring is a cold air flow path 3 and flows in a direction from the steam generator 9 to the reactor 1.
  • a main blower fan 10 is disposed above the housing of the steam generator 9.
  • the high temperature helium gas heated in the reactor 1 heats the secondary circuit water 8 in the steam generator 9, generating high temperature and high pressure steam 7, which is sent to the steam power system 13.
  • the steam generator 9 is a direct current steam generator with a spiral tube structure.
  • the reactor core 2 is designed as a flowable ball bed structure in which the spherical fuel element 33 flows from top to bottom.
  • the reactor core 2 may also be a fixedly arranged prismatic structure in which the fuel element 33 is located.
  • the all-ceramic coated particle spherical fuel element 33 is loaded from the core top 5, discharged from the discharge pipe of the core bottom 6, and the discharged fuel element 33 is burned one by one, and the discharge fuel has been reached.
  • the spent fuel element 33 is discharged from the stack for storage, and the fuel element that has not reached the discharge burnup is reloaded into the core 2 to effect multiple cycles of the fuel element.
  • FIG. 3 is a schematic view showing the structure of an embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • the system for providing steam to a steam power system includes: a nuclear steam supply system that sequentially connects first and last to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, An oxygen generator 26, a feed water pump 27, and a high pressure heater 28, wherein the nuclear steam supply system is as described above according to an embodiment of the present invention Nuclear steam supply system.
  • This embodiment of the invention is a steam direct power cycle. After the steam generated by the plurality of NSSS modules 11 is connected in parallel, the high pressure cylinder 21 and the low pressure cylinder 22 are successively operated to drive the generator 14. The wet steam that has completed the work is released into the condenser 23, passes through the condensate pump 24, passes through the low pressure heater 25, the deaerator 26, the feed water pump 27, and the high pressure heater 28, and is sent to the steam generator 9, Complete a heat cycle.
  • Fig. 4 is a schematic view showing the structure of another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • the system for providing steam to a steam power system includes a nuclear steam supply system that is connected in series to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, and a deaerator.
  • a feed water pump 27 and a high pressure heater 28, between the high pressure cylinder 21 and the low pressure cylinder 22, a steam reheater 15 and an intermediate pressure cylinder 29 are sequentially connected, wherein the nuclear steam supply system is as described above according to the present invention.
  • a nuclear steam supply system of the 'embodiment of the invention is a nuclear steam supply system that is connected in series to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, and a deaerator
  • This embodiment of the invention provides a reheat steam power generation cycle scheme for a dedicated reheat nuclear steam supply system module.
  • One or more reheat nuclear steam supply system modules 11 are provided, equipped with a steam reheater 15 to reheat the steam. After the steam generated by the plurality of NSSS modules 11 is connected in parallel, the steam is first introduced into the high-pressure cylinder 21, and the steam flowing out from the high-pressure cylinder 21 is heated into the dedicated steam reheater 15, and then enters the intermediate cylinder 29 and the low-pressure cylinder 22 successively. Do work, drive the generator 14.
  • the wet steam that has been subjected to work is released from the condenser 23, passes through the condensate pump 24, passes through the low pressure heater 25, the deaerator 26, the feed water pump 27, and the high pressure heater 28, and is sent to the steam generator 9 , complete a thermal cycle.
  • FIG. 5 is a schematic view showing the structure of still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • the system for providing steam to a steam power system includes a nuclear steam supply system that is connected in series to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, and a deaerator.
  • a feed water pump 27 and a high pressure heater 28, between the high pressure cylinder 21 and the low pressure cylinder 22, a steam reheater 15 and an intermediate pressure cylinder 29 are connected in sequence, and the outlet of the high pressure heater 28 and the steam are further Heater 15
  • the preliminary heating section is connected, and the inlet of the steam generator 9 is connected to the preliminary heating section of the steam reheater 15, wherein the nuclear steam supply system is the nuclear steam supply system according to the embodiment of the invention described above.
  • This embodiment of the invention is a modification of the previous embodiment. Specifically, one or more reheat nuclear steam supply system modules 11 are provided, and a steam reheater 15 is provided, in addition to reheating the steam, the feed water can be initially heated. The preliminary heated feed water is passed to the evaporation NSSS module 11 for further heating. After the steam generated by the plurality of NSSS modules 11 is connected in parallel, the steam is first introduced into the high-pressure cylinder 21, and the steam flowing out from the high-pressure cylinder 21 is heated into the dedicated steam reheater 15, and then enters the intermediate cylinder 29 and the low-pressure cylinder 22 successively. Do work, drive the generator 14.
  • the wet steam that has been subjected to work releases heat in the condenser 23, passes through the condensate pump 24, passes through the low pressure heater 25, the deaerator 26, the feed water pump 27, and the high pressure heater 28, and is sent to the steam reheater 15
  • the initial heating section completes a thermodynamic cycle.
  • Fig. 6 is a schematic view showing the structure of still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • the system for providing steam to a steam power system includes: a nuclear steam supply system that sequentially connects first and last to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, An oxygen generator 26, a feed water pump 27, and a high pressure heater 28, the outlet of the high pressure cylinder 21 being connected to a reheating portion of the steam generator 9, wherein the nuclear steam supply system is the nuclear steam according to an embodiment of the present invention described above Supply system.
  • This embodiment of the invention is an in-stack reheat steam power generation cycle scheme.
  • the high pressure cylinder 21 is firstly operated, and the steam flowing out from the high pressure cylinder 21 is again heated into the reheater portion of the steam generator 9 to be heated, and then enters the intermediate pressure cylinder 29 and the low pressure.
  • the cylinder 22 performs work to drive the generator 14.
  • the wet steam that has completed the work is released into the condenser 23, passes through the condensate pump 24, passes through the low pressure heater 25, the deaerator 26, the feed water pump 27, and the high pressure heater 28, and is sent to the steam generator 9, Complete a heat cycle.
  • FIG. 7 is a diagram showing still another embodiment of the high temperature gas cooled reactor steam power generation system of the present invention.
  • the system for providing steam to a steam power system includes: a nuclear steam supply system that sequentially connects first and last to form a closed steam circuit, a high pressure cylinder 21, a low pressure cylinder 22, a condenser 23, a condensate pump 24, a low pressure heater 25, An oxygen device 26, a feed water pump 27, and a high pressure heater 28 are connected to the reheater 30 and the intermediate pressure cylinder 29, respectively, and the outlet of the intermediate pressure cylinder 29 is connected to the reheater 30, the reheating
  • the compressor 30 is coupled to the low pressure cylinder 22, wherein the nuclear steam supply system is the nuclear steam supply system according to an embodiment of the present invention described above.
  • the reheater 15 is a helium-steam reheater and the reheater 30 is a steam-steam reheater.
  • This embodiment of the invention is an off-grid reheat steam power generation cycle scheme.
  • the high pressure cylinder 21 is firstly operated, and some of the steam flowing out of the high pressure cylinder 21 enters the intermediate pressure cylinder 29 for work, and the other portion enters the reheater 30 for the intermediate pressure cylinder 29.
  • the outlet steam is heated.
  • the heated steam re-enters the low pressure cylinder 22 to perform work to drive the generator 14.
  • the worked wet steam is released in the condenser 23, passes through the condensate pump 24, passes through the low pressure heater 25, the deaerator 26, the high pressure water pump 27, and the high pressure heater 28, and is sent to the steam generator 9, Complete a heat cycle.
  • the economic advantages of the modular ball-bed high-temperature gas-cooled reactor are mainly reflected in: 1) high core outlet temperature and high power generation efficiency; 2) non-stop stacking and unloading fuel, high power station availability; 3) no Requires emergency core cooling system, system simplification; 4) Modular manufacturing; 5) If a plurality of nuclear steam supply systems (NSSS) modules recommended in accordance with the present invention are used with a one-machine solution, the power scale of the unit can be further improved. Economic.
  • NSS nuclear steam supply systems
  • Modular high-temperature gas-cooled reactors The heat power of a single NSSS module is generally between 200-600 MW, and the corresponding electric power is generally more than 100,000 kW, while the electric power of a steam turbine generator set can be as high as one million kilowatts.
  • several steam turbine units are matched in parallel by several NSSS modules, that is, the "multi-NSSS module with one machine" is used to match the modular high-temperature gas-cooled reactor with the high-power steam unit.
  • the scale effect is realized by the form of batch copying of the NSSS module.
  • the high-temperature gas-cooled reactor can provide more than 90CTC heat source and can be coupled with supercritical steam power circulation technology to achieve power generation efficiency over other stack types. Even compared with conventional fossil fuel power plants of the same capacity, since the high-temperature gas-cooled reactor primary circuit is closed, there is no tail smoke loss, and it has the potential to achieve higher efficiency than the supercritical thermal power plant.
  • the invention forms a nuclear steam supply system (NSSS) module by using a reactor core and a pressure vessel with a steam generator as a standard module.
  • the NSSS modules have been replicated to provide steam for a large steam turbine power generation system, ie multiple NSSS modules with a steam turbine to achieve a "multi-NSSS module with one machine” configuration mode.
  • Smaller scale single modules reduce manufacturing complexity, and NSSS modules reduce cost due to bulk copying.
  • the NSSS module shares some auxiliary systems to improve the utilization of the auxiliary system and further reduce costs.
  • Multiple “multi-NSSS modules with one machine” generator sets can be configured in one site to further share power plant auxiliary facilities and reduce construction and operating costs.

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Description

高温气冷堆蒸汽发电系统及方法 技术领域
本发明涉及核电领域，特别是涉及一种高温气冷堆蒸汽发电系统 及方法。 背景技术
作为清洁、 安全、 环保的能源， 核能发电对于缓解能源安全和全 球气候变化问题都具有重要意义。虽然经过了美国三哩岛和前苏联切 尔诺贝利核电站事故挫折， 人们仍然在积极开发更安全、 经济性更好 的核能发电技术。 目前， 第三代核电技术巳经基本成熟。
正在研发的第四代核能系统中，高温气冷堆可以实现很高的出口 温度， 具有高发电效率和高品位热能供应能力， 从而引起人们广泛关 注。
高温气冷堆采用陶瓷型包覆颗粒燃料元件， 氦气作为冷却剂， 石 墨作为慢化剂， 堆芯出口温度可以达到 700°C至 950°C。 高温气冷堆 是一种安全性能好的堆型， 原因如下： 1 )优异的燃料元件性能； 2 ) 石墨堆芯的热容量大； 3 )全范围的负反应性温度系数； 4 )冷却剂氦 气为惰性气体， 化学稳定性好， 不会发生相变。
国际上高温气冷堆的发展起始于上世纪六十年代初， 在英国、德 国和美国相继建成了三座实验堆，到七十年代美国和德国分别建成和 运行了两座电功率为 330MW和 300MW的原型电站。 早期的高温气 冷堆在失去冷却剂的事故条件下， 若不采取专门措施， 堆芯最高温度 可能达到 2000°C以上， 因此需要专设应急堆芯冷却系统防止燃料元 件过热损坏。
为进一步提高反应堆的安全性， "模块式" 高温气冷堆的概念应 运而生。 模块式高温气冷堆特指具有固有安全特性、 单堆功率规模较 小的高温气冷堆。 这种反应堆的基本特点是： 在任何事故条件下反应 堆堆芯的剩余发热均能够通过非能动的方式载出，堆芯燃料最高温度 不会超过允许的限值。 由于避免了堆芯熔化的可能， 即使发生很低概 率的超设计基准事故， 核电厂厂外的放射性剂量仍在限值范围内， 技 术上可以不用釆取厂外应急计划。
根据燃料元件的形状不同， 高温气冷堆被分为球床堆与柱状堆。 前者把包覆颗粒燃料与石墨基体一起压制成直径 6厘米的燃料球，形 成能流动的球床堆芯， 实现不停堆在线更换核燃料。 后者把包覆颗粒 燃料与石墨一起压制成圆柱状芯块，然后再放入六角形棱柱形燃料组 件中， 形成固定型的棱柱状堆芯。
相对于柱状堆， 球床式高温气冷堆具有如下特点： 1 ) 不停堆装 卸燃料元件， 电站可用率高； 2 )堆芯过剩反应性小， 便于反应性控 制， 中子经济性高； 3 )卸料燃耗均匀， 卸料燃耗高， 燃料利用率高； 4 )正常运行时燃料颗粒温度低， 易于进一步提高反应堆出口温度。
作为面向发电上网的商业化电站， 不仅要有足够的安全性， 在经 济性方面也要体现出足够的竟争力。模块式高温气冷堆在经济性方面 的限制主要来自安全性考虑。模块式高温气冷堆的固有安全性要求事 故后衰变余热能够釆用非能动方式导出堆芯，保证燃料最高温度不超 过设计限值， 在技术上对单个堆芯的功率密度和总功率提出了限制。
如何在较小的单堆功率限制下实现更好的经济性，成为高温气冷 堆核电站设计和商业化推广过程中必须考虑的问题。
发明内容
本发明的目的是提供一种能够克服现有技术缺陷，在保证安全性 的同时实现经济性的高温气冷堆蒸汽发电系统及方法。
为达到上述目的，提供一种依照本发明实施方式的高温气冷堆蒸 汽发电系统， 其包括： 首尾顺次连接以形成闭合蒸汽回路的若干核蒸 汽供应系统、 高压缸、 低压缸、 凝汽器、 凝结水泵、 低压加热器、 除 氧器、 给水泵和高压加热器。 优选地，所述高温气冷堆蒸汽发电系统在所述高压缸和低压缸之 间依次连接蒸汽再热器和中压缸。
优选地，所述高压加热器的出口与所述蒸汽再热器的初步加热段 连接， 且所述蒸汽发生器的入口与所述蒸汽再热器的初步加热段连 接。
优选地， 所述高压缸的出口与蒸汽发生器的再加热部分连接。 优选地， 所述高压缸分别与再热器以及中压缸连接， 所述中压缸 的出口与再热器连接， 所述再热器与所述低压缸连接。
优选地 ,所述核蒸汽供应系统包括分别设置在两个压力容器内的 反应堆和蒸汽发生器， 所述反应堆和蒸汽发生器之间由热气导管相 连， 在所述蒸汽发生器的壳体上部设置有主氦风机。
优选地， 所述反应堆具有设计为可流动的球床结构的堆芯， 燃料 元件位于所述堆芯内并可由堆芯的顶部流向堆芯的底部。
优选地， 所述反应堆具有固定布置的棱柱状结构的堆芯， 燃料元 件位于所述堆芯内。
优选地， 所述燃料元件釆用全陶瓷型包覆颗粒燃料元件。
优选地，所述蒸汽发生器为直流式蒸汽发生器，釆用螺旋管结构。 优选地， 所述热气导管釆用环状结构， 外环为用于使氦气由蒸汽 发生器流向反应堆的冷氦气流道， 内环为用于使氦气由反应堆流向蒸 汽发生器的热氦气流道。
本发明还提供了一种高温气冷堆蒸汽发电方法， 其包括步骤：
51 , 利用若干个所述核蒸汽供应系统产生蒸汽；
52,将所述蒸汽并联后依次送入高压缸和低压缸做功以驱动发电
S3， 做完功的湿蒸汽进入凝汽器放热， 之后依次经过凝结水泵、 低压加热器、 除氧器、 给水泵和高压加热器进入蒸汽发生器， 完成一 个热力循环； S4, 重复执行步骤 Sl-S3。
优选地， 在步骤 S2中， 将蒸汽送入高压缸做功后， 从所述高压 缸流出来的蒸汽进入蒸汽再热器加热，之后先后进入中压缸和低压缸 功。
优选地， 在步骤 S3中， 做完功的湿蒸汽在进入蒸汽发生器之前 要进入到蒸汽再热器的初步加热段加热。
优选地， 在步骤 S2中， 将蒸汽送入高压缸做功后， 从所述高压 缸流出来的蒸汽进入蒸汽发生器的再加热部分进行加热，之后先后进 入中压缸和低庄缸做功。
优选地， 在步骤 S2中， 将蒸汽送入高压缸做功后， 从所述高压 缸流出来的一部分蒸汽进入再热器直接进行加热，从高压缸流出来的 另一部分蒸汽先进入中压缸做功，然后利用再热器对其出口蒸汽进行 加热，最后将直接加热以及先经中压缸做功然后再加热的蒸汽送入低 压缸做功。
上述技术方案具有如下优点： 以一个反应堆堆芯、压力容器配一 个蒸汽发生器为一个标准模块， 形成核蒸汽供应系统 (NSSS)模块。 NSSS模块经过复制， 共同为一台大型蒸汽轮机发电系统提供蒸汽， 即多个 NSSS模块配一个汽轮机， 实现 "多 NSSS模块带一机" 的配 置模式。 较小规模的单个模块可减少制造难度， NSSS模块由于批量 复制而降低了造价。 另外， NSSS模块共享一些辅助系统， 提高辅助 系统的利用率，进一步降低了成本。在一个厂址内还可以配置多个"多 NSSS模块带一机" 的发电机组， 进一步共享电厂辅助设施， 降低建 造和运营成本。 这样， 一方面保证反应堆的固有安全性， 利用这种固 有安全性对系统进行简化； 另一方面通过批量复制、 共享辅助系统、 规模效应， 保证汽机系统和整个电站其它系统的规模经济性。
附图说明
图 1是具有本发明实施方式的核蒸汽供应系统（NSSS )模块的蒸 汽发电系统；
图 2是依据本发明实施方式的核蒸汽供应系统（NSSS )模块的结 构示意图；
图 3是本发明的高温气冷堆蒸汽发电系统的一个实施例的结构示 意图；
图 4是本发明的高温气冷堆蒸汽发电系统的另一个实施例的结构 示意图；
图 5是本发明的高温气冷堆蒸汽发电系统的又一个实施例的结构 示意图；
图 6是本发明的高温气冷堆蒸汽发电系统的再一个实施例的结构 示意图；
图 7是本发明的高温气冷堆蒸汽发电系统的再一个实施例的结构 示意图。
其中， 1 : 反应堆； 2: 堆芯； 3: 冷氦气流道； 4: 热氦气流道； 5: 堆芯顶部； 6: 堆芯底部； 7: 高温高压蒸汽； 8: 二回路水； 9: 蒸汽发生器； 10: 主氦风机； 11 : NSSS模块； 12: 核电站辅助系统; 13: 蒸汽动力系统； 14: 发电机； 15: 蒸汽再热器； 21 : 高压缸； 21: 低压缸； 23: 凝汽器； 24: 凝结水泵； 25: 低压加热器； 26: 除氧器; 27: 给水泵； 28: 高压加热器； 29: 中压缸； 30: 再热器； 32: 热气 导管； 33: 燃料元件。
具体实施方式
下面结合附图和实施例，对本发明的具体实施方式作进一步详细 描述。 以下实施例用于说明本发明， 但不用来限制本发明的范围。
如图 1所示为具有本发明的核蒸汽供应系统（NSSS )模块的蒸汽 发电系统。 以一个反应堆堆芯、压力容器配一个蒸汽发生器为一个标 准模块， 形成核蒸汽供应系统（NSSS )模块。 多个 NSSS模块 11 , 共 享核电站辅助系统 12, —起为蒸汽动力系统 13提供蒸汽， 推动发电机 14进行发电。 核电站辅助系统 12主要包括： 燃料装卸与贮存系统、 一 回路压力泄放系统、 氦净化与氦辅助系统、 气体釆样与分析系统、 余 热排出系统、 蒸汽发生器事故排放系统、 设备冷却水系统、 反应堆厂 房通风和空调系统、 液体废物处理系统、 固体废物处理与存放系统、 核岛消防系统等。
图 2是依据本发明实施方式的核蒸汽供应系统（NSSS )模块的结 构示意图。 在 NSSS模块 11中， 反应堆 1与蒸汽发生器 9分别设置在两 个压力容器内， 其间用热气导管 32相连接， 构成 "肩并肩" 的布置方 式。 反应堆 1的压力容器、蒸汽发生器 9的壳体与热气导管 32的壳体组 成一回路压力边界， 并安装在一个混凝土屏蔽舱室内。 热气导管 32 采用环状结构， 内环为热氦气流道 4， 流动方向为从反应堆 1向蒸汽发 生器 9。 外环为冷氦气流道 3 , 流动方向为从蒸汽发生器 9到反应堆 1。 在蒸汽发生器 9的壳体上部布置了主氦风机 10。在反应堆 1中被加热的 高温氦气在蒸汽发生器 9中加热二回路水 8 , 产生高温高压蒸汽 7， 送 到蒸汽动力系统 13。 蒸汽发生器 9为直流式蒸汽发生器， 釆用螺旋管 结构。
反应堆堆芯 2设计为可流动的球床结构， 球形燃料元件 33自上向 下流动。 所述反应堆堆芯 2也可以为固定布置的棱柱状结构， 燃料元 件 33位于所述堆芯 2内。 釆用全陶瓷型包覆颗粒球形燃料元件 33 , 从 堆芯顶部 5装入， 从堆芯底部 6的卸料管卸出， 卸出的燃料元件 33逐个 进行燃耗测量， 已达到卸料燃耗的燃料元件 33被排出堆外贮存， 未达 到卸料燃耗的燃料元件则被重新装入堆芯 2,实现燃料元件多次循环。
图 3 是本发明的高温气冷堆蒸汽发电系统的一个实施例的结构 示意图。 所述为蒸汽动力系统提供蒸汽的系统包括： 首尾顺次连接以 形成闭合蒸汽回路的核蒸汽供应系统、 高压缸 21、 低压缸 22、 凝汽 器 23、 凝结水泵 24、 低压加热器 25、 除氧器 26、 给水泵 27和高压 加热器 28， 其中所述核蒸汽供应系统为上面所述根据本发明实施方 式的核蒸汽供应系统。
本发明的该实施例， 为蒸汽直接发电循环方案。 多个 NSSS模块 11产生的蒸汽经过并联后， 先后进入高压缸 21和低压缸 22做功， 驱动发电机 14。 做完功的湿蒸汽在凝汽器 23中放热， 经过凝结水泵 24, 再经过低压加热器 25、 除氧器 26、 给水泵 27和高压加热器 28 后， 被送入蒸汽发生器 9, 完成一个热力循环。
图 4是本发明的高温气冷堆蒸汽发电系统的另一个实施例的结 构示意图。所述为蒸汽动力系统提供蒸汽的系统包括首尾顺次连接以 形成闭合蒸汽回路的核蒸汽供应系统、 高压缸 21、 低压缸 22、 凝汽 器 23、 凝结水泵 24、 低压加热器 25、 除氧器 26、 给水泵 27和高压 加热器 28 , 在所述高压缸 21和低压缸 22之间依次连接蒸汽再热器 15和中压缸 29, 其中所述核蒸汽供应系统为上面所述根据本发明 '实 施方式的核蒸汽供应系统。
本发明的该实施例，为专用再热核蒸汽供应系统模块提供再热蒸 汽发电循环方案。 专门设置一个或多个再热核蒸汽供应系统模块 11， 配备蒸汽再热器 15 , 对蒸汽进行再热。 多个 NSSS模块 11产生的蒸汽 经过并联后， 先进入高压缸 21做功， 从高压缸 21流出的蒸汽， 进入专 设的蒸汽再热器 15得到加热， 然后先后进入中压缸 29和低压缸 22做 功， 驱动发电机 14。 做完功的湿蒸汽在凝 ·汽器 23中放热， 经过凝结水 泵 24, 再经过低压加热器 25、 除氧器 26、 给水泵 27和高压加热器 28 后， 被送入蒸汽发生器 9, 完成一个热力循环。
图 5 是本发明的高温气冷堆蒸汽发电系统的又一个实施例的结 构示意图。所述为蒸汽动力系统提供蒸汽的系统包括首尾顺次连接以 形成闭合蒸汽回路的核蒸汽供应系统、 高压缸 21、 低压缸 22、 凝汽 器 23、 凝结水泵 24、 低压加热器 25、 除氧器 26、 给水泵 27和高压 加热器 28, 在所述高压缸 21和低压缸 22之间依次连接蒸汽再热器 15和中压缸 29, 所述高压加热器 28的出口与所述蒸汽再热器 15的 初步加热段连接， 且所述蒸汽发生器 9的入口与所述蒸汽再热器 15 的初步加热段连接，其中所述核蒸汽供应系统为上面所述根据本发明 实施方式的核蒸汽供应系统。
本发明该实施例为上一个实施例的改进型。专门设置一个或多个 再热核蒸汽供应系统模块 11 , 配备的蒸汽再热器 15 , 除对蒸汽进行再 热外， 还可对给水进行初步加热。 得到初步加热后的给水进入蒸发 NSSS模块 11得到进一步加热。 多个 NSSS模块 11产生的蒸汽经过并联 后， 先进入高压缸 21做功， 从高压缸 21流出的蒸汽， 进入专设的蒸汽 再热器 15得到加热， 然后先后进入中压缸 29和低压缸 22做功， 驱动发 电机 14。 做完功的湿蒸汽在凝汽器 23中放热， 经过凝结水泵 24, 再经 过低压加热器 25、 除氧器 26、 给水泵 27和高压加热器 28后， 被送入蒸 汽再热器 15的初步加热段， 完成一个热力循环。
图 6 是本发明的高温气冷堆蒸汽发电系统的再一个实施例的结 构示意图。 所述为蒸汽动力系统提供蒸汽的系统包括： 首尾顺次连接 以形成闭合蒸汽回路的核蒸汽供应系统、 高压缸 21、 低压缸 22、 凝 汽器 23、 凝结水泵 24、 低压加热器 25、 除氧器 26、 给水泵 27和高 压加热器 28 , 所述高压缸 21的出口与蒸汽发生器 9的再加热部分连 接，其中所述核蒸汽供应系统为上面所述根据本发明实施方式的核蒸 汽供应系统。
本发明的该实施例， 为堆内再热蒸汽发电循环方案。 多个 NSSS 模块 11产生的蒸汽经过并联后， 先进入高压缸 21做功， 从高压缸 21 流出的蒸汽， 再次进入蒸汽发生器 9的再热器部分得到加热， 然后先 后进入中压缸 29和低压缸 22做功， 驱动发电机 14。 做完功的湿蒸汽在 凝汽器 23中放热，经过凝结水泵 24,再经过低压加热器 25、除氧器 26、 给水泵 27和高压加热器 28后， 被送入蒸汽发生器 9, 完成一个热力循 环。
图 7是本发明的高温气冷堆蒸汽发电系统的再一个实施例的结 构示意图。 所述为蒸汽动力系统提供蒸汽的系统包括： 首尾顺次连接 以形成闭合蒸汽回路的核蒸汽供应系统、 高压缸 21、 低压缸 22、 凝 汽器 23、 凝结水泵 24、 低压加热器 25、 除氧器 26、 给水泵 27和高 压加热器 28， 所述高压缸 21分别与再热器 30以及中压缸 29连接， 所述中压缸 29的出口与再热器 30连接， 所述再热器 30与所述低压 缸 22连接， 其中所述核蒸汽供应系统为上面所述根据本发明实施方 式的核蒸汽供应系统。 再热器 15为氦气-蒸汽再热器， 再热器 30为 蒸汽-蒸汽再热器。
本发明的该实施例为堆外再热蒸汽发电循环方案。 多个 NSSS模 块 11产生的蒸汽经过并联后， 先进入高压缸 21做功， 从高压缸 21流出 的蒸汽中， 一部分进入中压缸 29做功， 另一部分进入再热器 30, 对中 压缸 29的出口蒸汽进行加热。被加热的蒸汽再进入低压缸 22做功， 驱 动发电机 14。 作完功的湿蒸汽在凝汽器 23中放热， 经过凝结水泵 24， 再经过低压加热器 25、 除氧器 26、 高压水泵 27和高压加热器 28后， 被 送入蒸汽发生器 9, 完成一个热力循环。
模块式球床高温气冷堆在经济性方面的优势主要体现在： 1 ) 堆 芯出口温度高， 发电效率相应较高； 2 )不停堆装卸燃料， 较高的电 站可用率； 3 )不需要应急堆芯冷却系统， 系统简化； 4 )模块化制造; 5 )若釆用按照本发明所推荐的多个核蒸汽供应系统 (NSSS)模块带一 机方案， 提高机组功率规模， 可进一步提高经济性。
模块式高温气冷堆单个 NSSS模块热功率一般在 200-600MW之 间， 对应电功率一般为十几万千瓦， 而蒸汽透平发电机组的电功率可 高达百万千瓦级。 根据蒸汽透平的输入功率需要， 由数个 NSSS模块 并联匹配一台汽轮机组， 即釆用 "多 NSSS模块带一机"， 以实现模 块式高温气冷堆与大功率蒸汽机组的匹配。通过 NSSS模块批量复制 的形式， 实现规模效应。
充分利用高温气冷堆的 "高温" 特性， 实现超临界蒸汽循环， 提 高发电效率。高温气冷堆技术与目前已经普遍应用的蒸汽动力循环超 临界发电技术相结合， 是最可能实现的超临界循环核电站。 高温气冷 堆作为一个高质量的 "锅炉 " , 可以提供超过 90CTC的热源， 可与超 临界蒸汽动力循环技术耦合， 实现超过其它堆型的发电效率。 即使与 同容量的常规化石燃料电站相比， 由于高温气冷堆一回路是闭式的， 不存在尾烟损失， 有潜力实现比超临界火电站更高的效率。
以上所述仅是本发明的优选实施方式， 应当指出， 对于本技术领 域的普通技术人员来说， 在不脱离本发明技术原理的前提下， 还可以 做出若干改进和变型， 这些改进和变型也应视为本发明的保护范围。 工业实用性
本发明通过采用以一个反应堆堆芯、压力容器配一个蒸汽发生器 为一个标准模块， 形成核蒸汽供应系统（NSSS )模块。 NSSS模块经 过复制， 共同为一台大型蒸汽轮机发电系统提供蒸汽 , 即多个 NSSS 模块配一个汽轮机， 实现 "多 NSSS模块带一机" 的配置模式。 较小 规模的单个模块可减少制造难度， NSSS模块由于批量复制而降低了 造价。 另外， NSSS模块共享一些辅助系统， 提高辅助系统的利用率， 进一步降低了成本。 在一个厂址内还可以配置多个 "多 NSSS模块带 一机"的发电机组，进一步共享电厂辅助设施，降低建造和运营成本。 这样， 一方面保证反应堆的固有安全性， 利用这种固有安全性对系统 进行简化； 另一方面通过批量复制、 共享辅助系统、 规模效应， 保证 汽机系统和整个电站其它系统的规模经济性。

Claims

权 利 要 求
1、 一种高温气冷堆蒸汽发电系统， 其特征在于， 所述高温气冷堆蒸汽发 电系统包括： 首尾顺次连接以形成闭合蒸汽回路的若干核蒸汽供应系统、 高 压缸（21)、 低压缸（22)、 凝汽器（23)、 凝结水泵（24)、 低压加热器（25)、 除氧器（26)、 给水泵（27) 和高压加热器（28)。
2、 如权利要求 1所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述高 温气冷堆蒸汽发电系统在所述高压缸 （21) 和低压缸 （22) 之间依次连接蒸 汽再热器 (15) 和中压缸 （29)。
3、 如权利要求 2所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述高 压加热器（28) 的出口与所述蒸汽再热器（15) 的初步加热段连接， 且所述 蒸汽发生器（9) 的入口与所述蒸汽再热器 （15) 的初步加热段连接。
4、 如权利要求 1所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述高 压缸（21) 的出口与蒸汽发生器（9) 的再加热部分连接。
5、 如权利要求 1所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述髙 压缸（21 )分别与再热器 （30) 以及中压缸 （29)连接， 所述中压缸（29) 的出口与再热器（30)连接， 所述再热器（30) 与所述低压缸（22)连接。
6、如权利要求 1-5任一项所述的高温气冷堆蒸汽发电系统，其特征在于， 所述核蒸汽供应系统包括分别设置在两个压力容器内的反应堆（ 1 )和蒸汽发 生器（9), 所述反应堆（1)和蒸汽发生器（9)之间由热气导管 （32)相连, 在所述蒸汽发生器（9) 的壳体上部设置有主氦风机（10)。
7、 如权利要求 6所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述反 应堆（1)具有设计为可流动的球床结构的堆芯 （2), 燃料元件（33)位于所 述堆芯 （2) 内并可由堆芯 （2) 的顶部 （5)流向堆芯 （2) 的底部 （6)。
8、 如杈利要求 6所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述反 应堆（1)具有固定布置的棱柱状结构的堆芯 （2), 燃料元件（33)位于所述 堆芯 (2) 内。
9、 如杈利要求 7或 8所述的高温气冷堆蒸汽发电系统， 其特征在于， 所 述燃料元件（33 )釆用全陶瓷型包覆颗粒燃料元件。 ；
10、 如杈利要求 6所述的高温气冷堆蒸汽发电系统， 其特征在于 所述 蒸汽发生器（9) 为直流式蒸汽发生器， 釆用螺旋管结构。
11、 如权利要求 6所述的高温气冷堆蒸汽发电系统， 其特征在于， 所述 热气导管（32)釆用环状结构， 外环为用于使氦气由蒸汽发生器（9)流向反 应堆 （1) 的冷氦气流道（3), 内环为用于使氦气由反应堆 （1)流向蒸汽发 生器（9) 的热氦气流道（4)。
12、 一种高温气冷堆蒸汽发电方法， 其特征在于， 所述方法包括步骤：
51, 利用若干个所述核蒸汽供应系统产生蒸汽;
52, 将所述蒸汽并联后依次送入高压缸 (21) 和低压缸 (22)做功以驱 动发电机（14);
53,做完功的湿蒸汽进入凝汽器（ 23 )放热,之后依次经过凝结水泵（24)、 低压加热器（25)、 除氧器（26)、 给水泵（27) 和高压加热器（28)进入蒸 汽发生器（9), 完成一个热力循环；
54, 重复执行步骤 Sl-S3。
13、 如权利要求 12所述的高温气冷堆蒸汽发电方法， 其特征在于， 在步 骤 S2中， 将蒸汽送入高压缸（21)做功后， 从所述高压缸（21)流出来的蒸 汽进入蒸汽再热器（15)加热， 之后先后进入中压缸 （29) 和低压缸 （22) 做功。
14、 如杈利要求 13所述的高温气冷堆蒸汽发电方法， 其特征在于， 在步 骤 S3中， 做完功的湿蒸汽在进入蒸汽发生器（9)之前要进入到蒸汽再热器
(15) 的初步加热段加热。
15、 如权利要求 12所述的高温气冷堆蒸汽发电方法， 其特征在于， 在步 骤 S2中， 将蒸汽送入高压缸（21)做功后， 从所述高压缸（21)流出来的蒸 汽进入蒸汽发生器（9) 的再加热部分进行加热， 之后先后进入中压缸 （29) 和低压缸（22)做功。
16、 如杈利要求 12所述的高温气冷堆蒸汽发电方法， 其特征在于， 在步 骤 S2中， 将蒸汽送入高压缸（21)做功后， 从所述高压缸（21)流出来的一 部分蒸汽进入再热器（30)直接进行加热， 从高压缸（21 )流出来的另一部 分蒸汽先进入中压缸 （29)做功， 然后利用再热器（30) 对其出口蒸汽进行 加热， 最后将直接加热以及先经中压缸做功然后再加热的蒸汽送入低压缸 (22)做功。
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