WO2016078285A1 - 二次侧非能动佘热导出系统 - Google Patents

二次侧非能动佘热导出系统 Download PDF

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Publication number
WO2016078285A1
WO2016078285A1 PCT/CN2015/075498 CN2015075498W WO2016078285A1 WO 2016078285 A1 WO2016078285 A1 WO 2016078285A1 CN 2015075498 W CN2015075498 W CN 2015075498W WO 2016078285 A1 WO2016078285 A1 WO 2016078285A1
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WIPO (PCT)
Prior art keywords
steam
cooling water
water tank
steam generator
secondary side
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PCT/CN2015/075498
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English (en)
French (fr)
Inventor
沈永刚
卢向晖
周洲
芮旻
张吉胜
侯华青
崔旭阳
林支康
Original Assignee
中科华核电技术研究院有限公司
中国广核集团有限公司
中国广核电力股份有限公司
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Application filed by 中科华核电技术研究院有限公司, 中国广核集团有限公司, 中国广核电力股份有限公司 filed Critical 中科华核电技术研究院有限公司
Priority to GB1600378.2A priority Critical patent/GB2535848B/en
Publication of WO2016078285A1 publication Critical patent/WO2016078285A1/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
    • G21C15/18—Emergency cooling arrangements; Removing shut-down heat
    • 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
    • G21C9/00—Emergency protection arrangements structurally associated with the reactor, e.g. safety valves provided with pressure equalisation devices
    • 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 invention relates to the field of nuclear power plant safety equipment, in particular to a secondary side passive heat extraction system for a pressurized water reactor nuclear power plant with a one-pass type steam generator.
  • nuclear power is a major breakthrough in the history of energy use.
  • the fission reaction of the atomic nucleus can produce high-energy output that is unmatched by all other traditional fossil energy sources. These high-energy outputs often require only a small amount of nuclear fuel.
  • the characteristics of low input and high output have made centuries pay more attention to the use of nuclear energy and continue to increase research and development in the field of nuclear energy.
  • nuclear energy has become an important energy component of many countries in the world.
  • nuclear power has a very high use value, and it may also cause great harm.
  • In the process of using nuclear power if there is a major accident such as a nuclear leak caused by improper protection, it will affect the environment around the nuclear power plant and even the whole civilization. Bringing extremely serious nuclear pollution disasters.
  • this kind of hot exhaust system requires power supply, emergency equipment (such as expensive emergency diesel engine) and operator intervention outside the plant under accident conditions, which increases the risk of operator error.
  • emergency equipment such as expensive emergency diesel engine
  • operator intervention outside the plant under accident conditions which increases the risk of operator error.
  • the number of equipment has been greatly increased, thereby increasing the cost of equipment purchase, installation, operation and maintenance, and correspondingly increasing the construction cost and operation and maintenance cost of the nuclear power plant.
  • the technical solution of the present invention is to provide a secondary side passive heat extraction system for deriving the core decay heat in the containment, which includes a steam line and a water supply line, the steam a pipeline sealingly penetrating through the containment vessel and connected to an outlet of a steam generator disposed in the containment vessel and a cooling water tank disposed outside the containment vessel, the water supply pipeline sealingly penetrating through the containment vessel and connected to the An inlet of the cooling water tank and the steam generator, the steam line, the water supply line, and the cooling water tank form a circulation passage to discharge decay heat in the containment out of the containment.
  • the location of the cooling water tank is higher than the position of the steam generator.
  • the inlet end of the steam line is connected to the outlet of the steam generator, and the outlet end of the steam line extends below the level of the cooling water in the cooling water tank.
  • the inlet end of the water supply line is connected to the bottom of the cooling water tank, and the outlet end of the water supply line is connected to the inlet of the steam generator.
  • the steam line is provided with a first valve, and the first valve is located in the safety enclosure.
  • the water supply line is provided with a second valve and a third valve, the second valve is located outside the safety enclosure, and the third valve is located in the safety enclosure.
  • the outlet of the steam generator is at the upper end and the inlet of the steam generator is at the lower end.
  • the steam generator is connected to a reactor pressure vessel in the containment, and the steam generator is further connected to the main feed water line and the main steam line, respectively.
  • a fourth valve is disposed on the main water supply line, and a fifth valve is disposed on the main steam line, and the fourth valve and the fifth valve are all located in the safety shell.
  • the cooling water tank is disposed in an open position.
  • the steam generator is a one-pass type steam generator.
  • the secondary side passive heat transfer system of the present invention includes a steam line and a water supply line, and the steam line is sealingly penetrates the safety shell and is connected to steam generated in the safety shell.
  • An outlet of the device and a cooling water tank disposed outside the containment the water supply line sealingly penetrating the a containment vessel coupled to the cooling water tank and an inlet of the steam generator, the steam line, the water supply line, the cooling water tank forming a circulation passage to derive decay heat within the containment vessel from the containment vessel outer.
  • the steam in the steam generator enters the cooling water tank through the steam line, is condensed in the cooling water tank, and the water in the cooling water tank flows back to the steam generator through the water supply line, and the circulation passage forms a passive natural circulation by the density difference.
  • the circuit discharges the steam into the atmosphere by heating the cooling water in the cooling water tank. When the cooling water in the cooling water tank is consumed, the steam is directly discharged into the atmosphere, so that the core decay heat can be completely passive after the accident.
  • the discharge greatly reduces the possibility of system failure, and also avoids the safety of nuclear power plants due to the loss of power supply and operator error in the factory; no need for emergency equipment, thus greatly reducing equipment
  • the quantity reduce the cost of equipment purchase, installation, operation and maintenance, and reduce the construction cost and operation and maintenance cost of the nuclear power plant.
  • the steam line and the water supply line are directly connected to the steam generator on the secondary side, thereby avoiding the occurrence of steam generator drying and coolant leakage under the accident, and the cooling water tank is disposed outside the safety shell, saving the use of the safety shell. Space, and the use of open circuit, does not need to set the heat exchanger in the cooling water tank required by the traditional passive heat exhaust system, reducing the design and construction cost of the system.
  • FIG. 1 is a schematic view showing the structure of a secondary side passive heat extraction system of the present invention.
  • FIG. 2 is a schematic view showing the state of use of the secondary side passive heat transfer system of the present invention.
  • Fig. 3 is a schematic view showing another use state of the secondary side passive heat transfer system of the present invention.
  • the secondary side passive heat extraction system 100 provided by the present invention completely realizes the derivation of the core decay heat in the containment 110 under the accident, and can reduce the construction and operation and maintenance costs.
  • the safety vessel 110 is provided with a connected reactor pressure vessel 120 and a steam generator 130.
  • the pressurized water reactor nuclear power plant is designed with two to four steam generators 130, and only one of them is illustrated in this embodiment. The setting of the remaining steam generators 130 is well known to those skilled in the art.
  • steam generator 130 It is the only heat exchange device in this system and is used for heat transfer in the first loop.
  • the steam generator 130 is a one-pass type steam generator
  • the one-pass type steam generator (OTSG) is one of the foundations of the present invention, since the water volume of the one-pass type steam generator is small, The water in the cooling water tank 150 (described later) is easily filled with a single-pass steam generator to initially form a single-phase natural circulation; if a conventional steam generator is used, this capability is not available.
  • the secondary side passive heat extraction system 100 may be provided with only one set, and the set of secondary side passive heat extraction system 100 respectively corresponds to the plurality of steam generators 130; of course, multiple groups may also be provided.
  • the secondary side passive heat extraction system 100, each set of secondary side passive heat extraction system 100 corresponds to a steam generator 130.
  • a set of secondary side passive heat extraction system 100 corresponding to one steam generator 130 will be taken as an example to describe its structure.
  • the secondary side passive heat extraction system 100 includes a steam line 140 and a water supply line 160 , the steam line 140 sealingly penetrating through the containment 110 and connected to the safety housing 110 .
  • the steam line 140, the feed water line 160, and the cooling water tank 150 form a circulation passage to discharge the decay heat in the containment 110 out of the containment vessel 110.
  • the cooling water tank 150 is disposed in an open position, and the position of the cooling water tank 150 is higher than the position of the steam generator 130.
  • the cooling water tank 150 is provided with a quantity of water required to take away the core decay heat within a certain time after the accident.
  • the high position of the cooling water tank 150 is arranged to form a high level difference required for the natural circulation with the steam generator 130. Therefore, after the steam line 140 and the water supply line 160 are connected, the cooling water in the cooling water tank 150 automatically enters the steam generation.
  • the heater 130 is heated to take away the heat of the primary circuit; since the cooling water tank 150 is open, when the cooling water in the 150 is evaporated, the steam is directly discharged to the atmosphere via the steam line 140, and there is no need to add water to the cooling water tank 150. .
  • the cooling water tank 150 is disposed outside the safety enclosure 110 without occupying the space inside the safety enclosure 110, thereby saving the available space in the safety enclosure 110.
  • the outlet of the steam generator 130 is located at the upper end, the inlet of the steam generator 130 is located at the lower end, and the inlet end 140a of the steam line 140 is connected to the outlet of the steam generator 130, the steam line 140
  • the outlet end 140b extends from above the cooling water tank 150, and the outlet end 140b of the steam line 140 extends to a depth below the level of the cooling water in the cooling water tank 150;
  • the inlet end 160a of the water supply line 160 is connected to At the bottom of the cooling water tank 150, the outlet end 160b of the water supply line 160 is connected to the inlet of the steam generator 130;
  • the steam generator 130, the steam line 140, the cooling water tank 150, and the water supply line 160 form a circulation passage of the open circuit Since the steam line 140 and the water supply line 160 are directly connected to the steam generator 130 on the secondary side, the occurrence of the steam generator 130 being dried and the primary coolant leaking under the accident is avoided.
  • a first valve 141 is disposed on the steam line 140, and the first valve 141 is located in the safety shell 110.
  • the water supply line 160 is provided with a second valve 161 and a third valve 162.
  • the second valve 161 is located outside the containment vessel 110, and the third valve 162 is located within the containment vessel 110.
  • one end of the steam generator 130 is also connected to the main water supply line 170, and the outlet of the steam generator 130 is also connected to the main steam line 180.
  • the primary reactor core When the nuclear power plant is in normal operation, the primary reactor core generates huge thermal energy due to fission of nuclear fuel, and the heat of the primary circuit is used to heat the feed water to generate steam.
  • the heat energy When the steam passes through the heat transfer tube in the steam generator 130, the heat energy is transmitted through the pipe wall.
  • the second-circuit cooling water outside the heat pipe, the heat-releasing feed water is sent back to the core by the main pump to be reheated and then enters the steam generator 130.
  • the secondary circuit cooling water is heated to become steam, and the steam enters the steam turbine through the main steam line 180 to perform work, thereby converting the heat energy into electric power; after the completion of the work, the steam enters the condenser to be cooled, and then condenses into water and then returns to the steam through the main water supply line 170.
  • Generator 130 is reheated into steam.
  • the main water supply line 170 is provided with a fourth valve 171, and the main steam line 180 is provided with a fifth valve 181, and the fourth valve 171 and the fifth valve 181 are all located in the safety shell 110, the nuclear power plant In normal operation, the fourth valve 171 and the fifth valve 181 are in an open state.
  • the secondary side passive heat extraction system 100 does not start, but is in an available state.
  • the first valve 141, the second valve 161 and the third valve The 162 is in the closed state, and the fourth valve 171 and the fifth valve 181 are in an open state.
  • the reactor under the accident condition (design basis accident to the over-design basis accident such as the whole plant power failure), the reactor is shut down, and the corresponding protection signal triggers the secondary side passive heat extraction system 100 to make It starts up.
  • the fourth valve 171 and the fifth valve 181 are isolated, and the first valve 141, the second valve 161, and the third valve 162 are sequentially opened, and the steam in the steam generator 130 enters the steam line 140 through the outlet thereof, and then passes through the steam.
  • the outlet end 140b of the line 140 enters the cooling water tank 150 where it is condensed, and the water in the cooling water tank 150 flows back to the steam generator 130 through the feed water line 160, which forms a passive natural circulation by density difference. Loop.
  • the heat is transferred to the cooling water in the cooling water tank 150.
  • the temperature of the cooling water in the cooling water tank 150 is continuously increased due to the heating, and the cooling water tank 150 is heated by heating.
  • the water is vented to the atmosphere by cooling water, and boiling occurs after reaching 100 ° C, and the water level in the cooling water tank 150 starts to gradually decrease.
  • the invention connects the steam line 140 and the water supply line 160 directly to the steam generator 130 on the secondary side and forms an open loop circuit, which can completely realize the discharge of the core decay heat after the accident, and greatly reduces the heat. The possibility of system failure.
  • the secondary side passive heat extraction system 100 of the present invention includes a steam line 140 and a water supply line 160
  • the steam line 140 sealingly penetrates the containment 110 and is connected to the steam generator 130 disposed in the containment vessel 110.
  • the water supply line 160 and the cooling water tank 150 form a circulation passage to guide the decay heat in the containment 110 out of the containment 110.
  • the steam in the steam generator 130 enters the cooling water tank 150 via the steam line 140, is condensed in the cooling water tank 150, and the cooling water in the cooling water tank 150 flows back to the steam generator 130 through the water supply line 160, the circulation passage Relying on the density difference to form a passive natural circulation loop, cooling the cooling tank 150 by heating
  • the water is discharged into the atmosphere, and when the cooling water in the cooling water tank 150 is consumed, the steam is directly discharged into the atmosphere, so that the core decay heat can be completely passively discharged after the accident, which is greatly reduced.
  • the possibility of system failure also avoids the safety of nuclear power plants due to the loss of power supply and operator error in the factory; no need for emergency equipment, thus greatly reducing the number of equipment, reducing equipment purchase, installation, The cost of operation and maintenance will reduce the construction cost and operation and maintenance cost of the nuclear power plant accordingly.
  • the steam line 140 and the water supply line 160 are directly connected to the steam generator 130 on the secondary side, thereby avoiding the occurrence of the steam generator 130 being dried and the coolant leaking under the accident; and the cooling water tank 150 is disposed outside the containment 110.
  • the available space of the containment vessel 110 is saved; and the open circuit is used, and the heat exchanger in the cooling water tank 150 required by the conventional passive heat exhaust system is not required, which reduces the design and construction cost of the system.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Structure Of Emergency Protection For Nuclear Reactors (AREA)

Abstract

一种二次侧非能动余热导出系统，包括蒸汽管线(140)及给水管线(160)，蒸汽管线(140)密封地贯穿安全壳并连接于设于安全壳内的蒸汽发生器(130)的出口及设于安全壳外的冷却水箱(150)，给水管线(160)密封地贯穿安全壳并连接于冷却水箱(150)及蒸汽发生器(130)的入口，蒸汽管线(140)、给水管线(160)、冷却水箱(150)形成循环通道以将安全壳内的衰变热导出安全壳外。蒸汽管线(140)、给水管线(160)直接与二次侧的蒸汽发生器(130)相连接，避免了事故下蒸汽发生器(130)烧干及冷却剂泄漏的情况出现，冷却水箱(150)设置于安全壳外，节约了安全壳的可用空间，且利用开式回路，不需要设置传统的非能动余热排出系统所需要的冷却水箱(150)中的换热器，不需要应急设备，从而降低了系统的设计和建造费用。

Description

二次侧非能动佘热导出系统 技术领域
本发明涉及核电站安全设备领域，尤其涉及一种用于具有一次通过型蒸汽发生器的压水堆核电厂的二次侧非能动佘热导出系统。
背景技术
核电的使用是人类在能源利用史上的一个重大突破，利用原子核的裂变反应，能够产生其他所有传统化石能源所无法比拟的高能量输出，并且这些高能量输出往往只需要耗费少量的核燃料，这种低投入高产出的特性，使得人类日益重视对核能的利用，并不断加大在核能领域的研究开发，时至今日，核能已成为世界上许多国家的重要能源组成部分。然而，核电具有极高利用价值的同时，也可能带来很大的危害，在利用核电的过程中，如果保护不当而致使出现核泄漏等重大事故，将会对核电厂周边的环境乃至全人类带来极其严重的核污染灾害。
核电站中，安全壳是反应堆的重要安全设施，是防止放射性产物释放到大气环境中的最后一道屏障。在现役的压水堆核电站中，为保证反应堆在发生设计基准事故和超设计基准事故时，堆芯的衰变热都能够被持续排出，通常在一次侧设计佘热排出系统，这种方式将冷却水箱设置于安全壳内，占用了安全壳的可用空间，而且还需要新增放置于换料水箱中的换热器，从而使系统的设计和建造费用增加，一旦发生传热管破损，还会使一回路冷却剂泄漏。此外，这种佘热排出系统在事故工况下，需要厂内厂外的电源、应急设备(例如昂贵的应急柴油机)及操纵员的干预，这一方面增大了操纵员人因失误的风险，另一方面大大增加了设备数量，由此增加设备购买、安装、运行和维修等的费用，相应增加核电厂的建造成本和运维费用。
因此，有必要提供一种完全非能动地实现事故下堆芯衰变热的导出、降低建造及运维成本的二次侧非能动佘热导出系统，以解决上述现有技术的不足。
发明内容
本发明的目的在于提供一种完全非能动地实现事故下堆芯衰变热的导出、降低建造及运维成本的二次侧非能动佘热导出系统。
为实现上述目的，本发明的技术方案为：提供一种二次侧非能动佘热导出系统，用于对安全壳内的堆芯衰变热进行导出，其包括蒸汽管线及给水管线，所述蒸汽管线密封地贯穿安全壳并连接于设于所述安全壳内的蒸汽发生器的出口及设于所述安全壳外的冷却水箱，所述给水管线密封地贯穿所述安全壳并连接于所述冷却水箱及所述蒸汽发生器的入口，所述蒸汽管线、所述给水管线、所述冷却水箱形成循环通道以将所述安全壳内的衰变热导出所述安全壳外。
较佳地，所述冷却水箱的位置高于所述蒸汽发生器的位置。
较佳地，所述蒸汽管线的入口端连接于所述蒸汽发生器的出口，所述蒸汽管线的出口端伸入所述冷却水箱内的冷却水的液面以下。
较佳地，所述给水管线的入口端连接于所述冷却水箱的底部，所述给水管线的出口端连接于所述蒸汽发生器的入口。
较佳地，所述蒸汽管线上设有第一阀门，所述第一阀门位于所述安全壳内。
较佳地，所述给水管线上设有第二阀门及第三阀门，所述第二阀门位于所述安全壳外，所述第三阀门位于所述安全壳内。
较佳地，所述蒸汽发生器的出口位于上端，所述蒸汽发生器的入口位于下端。
较佳地，所述蒸汽发生器与所述安全壳内的反应堆压力容器相连接，且所述蒸汽发生器还分别连接主给水管线及主蒸汽管线。
较佳地，所述主给水管线上设有第四阀门，所述主蒸汽管线上设有第五阀门，所述第四阀门、所述第五阀门均位于所述安全壳内。
较佳地，所述冷却水箱呈敞口设置。
较佳地，所述蒸汽发生器为一次通过型蒸汽发生器。
与现有技术相比，由于本发明的二次侧非能动佘热导出系统，包括蒸汽管线及给水管线，所述蒸汽管线密封地贯穿所述安全壳并连接于设于安全壳内的蒸汽发生器的出口及设于安全壳外的冷却水箱，所述给水管线密封地贯穿所述 安全壳并连接于所述冷却水箱及所述蒸汽发生器的入口，所述蒸汽管线、所述给水管线、所述冷却水箱形成循环通道以将所述安全壳内的衰变热导出所述安全壳外。事故时，蒸汽发生器内的蒸汽经蒸汽管线进入冷却水箱，在冷却水箱中被冷凝，冷却水箱中的水经过给水管线流回蒸汽发生器，所述循环通道依靠密度差形成非能动的自然循环回路，通过加热冷却水箱中的冷却水而将蒸汽排放到大气中，当冷却水箱中的冷却水消耗完后，蒸汽直接排放到大气中，因此可以在事故后完全非能动的实现堆芯衰变热的排出，极大的减小了系统失效的可能性，也避免了因为丧失厂内外电源和操纵员人因失误的危害，提高了核电厂的安全性；不需要应急设备，从而大大减少了设备数量，减少设备购买、安装、运行和维修等费用，相应减少核电厂的建造成本和运维费用。另外，蒸汽管线、给水管线直接与二次侧的蒸汽发生器相连接，避免了事故下蒸汽发生器烧干及冷却剂泄漏的情况出现，冷却水箱设置于安全壳外，节约了安全壳的可用空间，且利用开式回路，不需要设置传统的非能动佘热排出系统所需要的冷却水箱中的换热器，降低了系统的设计和建造费用。
附图说明
图1是本发明二次侧非能动佘热导出系统的结构示意图。
图2是本发明二次侧非能动佘热导出系统的使用状态示意图。
图3是本发明二次侧非能动佘热导出系统的另一使用状态示意图。
具体实施方式
现在参考附图描述本发明的实施例，附图中类似的元件标号代表类似的元件。
如图1所示，本发明所提供的二次侧非能动佘热导出系统100，完全非能动地实现事故下安全壳110内堆芯衰变热的导出，且能降低建造及运维成本。所述安全壳110内设有相连接的反应堆压力容器120及蒸汽发生器130，一般而言，压水堆核电厂设计有二至四台蒸汽发生器130，本实施例中仅示意出其中一台，其余蒸汽发生器130的设置为本领域技术人员所熟知的技术。且蒸汽发生器130 是本系统中唯一的换热设备，用于一回路热量导出。
本发明中，所述蒸汽发生器130为一次通过型蒸汽发生器，一次通过型蒸汽发生器(OTSG)是本发明的基础之一，由于一次通过型蒸汽发生器的水装量较小，因此，冷却水箱150(详见后述)的水很容易充满一次通过型蒸汽发生器，初始形成单相的自然循环；若采用传统的蒸汽发生器，则不具备此能力。
其中，所述二次侧非能动佘热导出系统100可仅设置一组，该组二次侧非能动佘热导出系统100分别与多个蒸汽发生器130相对应；当然，也可以设置多组二次侧非能动佘热导出系统100，每一组二次侧非能动佘热导出系统100与一蒸汽发生器130相对应。
下面继续参阅图1所示，以对应一个蒸汽发生器130设置的一组二次侧非能动佘热导出系统100为例，对其结构进行说明。
如图1所示，所述二次侧非能动佘热导出系统100包括蒸汽管线140及给水管线160，所述蒸汽管线140密封地贯穿所述安全壳110并连接于设于安全壳110内的蒸汽发生器130的出口及设于安全壳110外的冷却水箱150，所述给水管线160密封地贯穿所述安全壳110并连接于所述冷却水箱150及所述蒸汽发生器130的入口，所述蒸汽管线140、给水管线160、冷却水箱150形成循环通道以将所述安全壳110内的衰变热导出所述安全壳110外。
具体地，所述冷却水箱150呈敞口设置，且冷却水箱150的位置高于蒸汽发生器130的位置，冷却水箱150内装有能够在事故后一定时间内带走堆芯衰变热所需的水量，通过冷却水箱150的高位布置，使其与蒸汽发生器130形成自然循环所需的高位差，因此，在蒸汽管线140和给水管线160连通后，冷却水箱150中的冷却水会自动进入蒸汽发生器130被加热，带走一回路热量；由于冷却水箱150呈敞口设置，因此当150内的冷却水蒸发完时，蒸汽经蒸汽管线140直接排放到大气中，不需要向冷却水箱150内加水。而将冷却水箱150设于安全壳110外，不用占据安全壳110内的空间，从而节约安全壳110内的可用空间。
其中，所述蒸汽发生器130的出口位于上端，蒸汽发生器130的入口位于下端，蒸汽管线140的入口端140a连接于蒸汽发生器130的出口，蒸汽管线140 的出口端140b从冷却水箱150的上方伸入其内，且蒸汽管线140的出口端140b伸入冷却水箱150内的冷却水的液面以下一定深度；所述给水管线160的入口端160a连接于冷却水箱150的底部，所述给水管线160的出口端160b连接于所述蒸汽发生器130的入口；蒸汽发生器130、，蒸汽管线140、冷却水箱150、给水管线160形成开式回路的循环通道，由于蒸汽管线140、给水管线160直接与二次侧的蒸汽发生器130相连接，避免了事故下蒸汽发生器130烧干及一回路冷却剂泄漏的情况出现。
另外，所述蒸汽管线140上设有第一阀门141，所述第一阀门141位于所述安全壳110内。所述给水管线160上设有第二阀门161及第三阀门162，所述第二阀门161位于安全壳110外，第三阀门162位于安全壳110内。核电厂正常运行时，第一阀门141、第二阀门161、第三阀门162均关闭；事故后，打开第一阀门141、第二阀门161及第三阀门162以启动所述二次侧非能动佘热导出系统100。
本发明，蒸汽发生器130的一端还与主给水管线170连接，蒸汽发生器130的出口还与主蒸汽管线180连接。核电站正常运行时，一回路反应堆堆芯因核燃料裂变产生巨大的热能，利用一回路的热量加热给水而产生蒸汽，蒸汽通过蒸汽发生器130内的传热管时，通过管壁将热能传递给传热管外的二回路冷却水，释放热量的给水又被主泵送回堆芯重新加热再进入蒸汽发生器130。二回路冷却水受热从而变成蒸汽，蒸汽通过主蒸汽管线180进入汽轮机做功，从而把热能转化为电力；做完功后的蒸汽进入冷凝器冷却，凝结成水后再经主给水管线170返回蒸汽发生器130，重新加热成蒸汽。
其中，所述主给水管线170上设有第四阀门171，所述主蒸汽管线180上设有第五阀门181，所述第四阀门171、第五阀门181均位于安全壳110内，核电厂正常运行时，第四阀门171、第五阀门181处于打开状态。
下面结合图1-图3所示，对本发明二次侧非能动佘热导出系统100的工作原理进行说明。
如图1所示，核电厂正常运行的情况下，所述二次侧非能动佘热导出系统100不启动，但处于可用状态，此时，第一阀门141、第二阀门161及第三阀门 162处于关闭状态，第四阀门171、第五阀门181处于打开状态。
如图2所示，在事故工况(设计基准事故至全厂断电等超设计基准事故)下，反应堆停堆，相应的保护信号触发所述二次侧非能动佘热导出系统100而使其启动。此时，第四阀门171、第五阀门181被隔离，第一阀门141、第二阀门161、第三阀门162依次打开，蒸汽发生器130内的蒸汽通过其出口进入蒸汽管线140，再经蒸汽管线140的出口端140b进入冷却水箱150，在冷却水箱150中被冷凝，冷却水箱150中的水经过给水管线160流回蒸汽发生器130，所述循环通道内依靠密度差形成非能动的自然循环回路。而蒸汽在冷却水箱150中冷凝传热时，将热量传递给冷却水箱150中的冷却水，在事故早期，冷却水箱150中的冷却水的温度由于加热持续升高，通过加热冷却水箱150中的冷却水而将蒸汽排放到大气中，到达100℃后发生沸腾，冷却水箱150中的水位开始逐渐下降。
如图3所示，当冷却水箱150中的冷却水消耗一定时间后，蒸汽管线140的出口端140b裸露于空气中，蒸汽直接排放到大气中，热量不再进入冷却水箱150，但冷却水箱150中的水可持续消耗直至排空。当冷却水箱150中的冷却水消耗完一段时间后，若不向冷却水箱150补水，则系统功能丧失。
本发明将蒸汽管线140、给水管线160直接与二次侧的蒸汽发生器130相连接并形成开式循环回路，可以在事故后完全非能动的实现堆芯衰变热的排出，极大的减小了系统失效的可能性。
由于本发明二次侧非能动佘热导出系统100，包括蒸汽管线140及给水管线160，所述蒸汽管线140密封地贯穿所述安全壳110并连接于设于安全壳110内的蒸汽发生器130的出口及设于安全壳110外的冷却水箱150，所述给水管线160密封地贯穿所述安全壳110并连接于所述冷却水箱150及所述蒸汽发生器130的入口，所述蒸汽管线140、所述给水管线160、所述冷却水箱150形成循环通道以将所述安全壳110内的衰变热导出所述安全壳110外。事故时，蒸汽发生器130内的蒸汽经蒸汽管线140进入冷却水箱150，在冷却水箱150中被冷凝，冷却水箱150中的冷却水经过给水管线160流回到蒸汽发生器130，所述循环通道内依靠密度差形成非能动的自然循环回路，通过加热冷却水箱150中的冷却 水而将蒸汽排放到大气中，当冷却水箱150中的冷却水消耗完后，蒸汽直接排放到大气中，因此可以在事故后完全非能动的实现堆芯衰变热的排出，极大的减小了系统失效的可能性，也避免了因为丧失厂内外电源和操纵员人因失误的危害，提高了核电厂的安全性；不需要应急设备，从而大大减少了设备数量，减少设备购买、安装、运行和维修等费用，相应减少核电厂的建造成本和运维费用。另外，蒸汽管线140、给水管线160直接与二次侧的蒸汽发生器130相连接，避免了事故下蒸汽发生器130烧干及冷却剂泄漏的情况出现；而冷却水箱150设置于安全壳110外，节约了安全壳110的可用空间；且利用开式回路，不需要设置传统的非能动佘热排出系统所需要的冷却水箱150中的换热器，降低了系统的设计和建造费用。
本发明中反应堆压力容器120及蒸汽发生器130的设置，为本领域普通技术人员所熟知，在此不再做详细的说明。
以上所揭露的仅为本发明的优选实施例而已，当然不能以此来限定本发明之权利范围，因此依本发明申请专利范围所作的等同变化，仍属本发明所涵盖的范围。

Claims (11)

  1. 一种二次侧非能动余热导出系统，用于对安全壳内的堆芯衰变热进行导出，其特征在于：包括蒸汽管线及给水管线，所述蒸汽管线密封地贯穿安全壳并连接于设于所述安全壳内的蒸汽发生器的出口及设于所述安全壳外的冷却水箱，所述给水管线密封地贯穿所述安全壳并连接于所述冷却水箱及所述蒸汽发生器的入口，所述蒸汽管线、所述给水管线、所述冷却水箱形成循环通道以将所述安全壳内的衰变热导出所述安全壳外。
  2. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述冷却水箱的位置高于所述蒸汽发生器的位置。
  3. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述蒸汽管线的入口端连接于所述蒸汽发生器的出口，所述蒸汽管线的出口端伸入所述冷却水箱内的冷却水的液面以下。
  4. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述给水管线的入口端连接于所述冷却水箱的底部，所述给水管线的出口端连接于所述蒸汽发生器的入口。
  5. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述蒸汽管线上设有第一阀门，所述第一阀门位于所述安全壳内。
  6. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述给水管线上设有第二阀门及第三阀门，所述第二阀门位于所述安全壳外，所述第三阀门位于所述安全壳内。
  7. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述蒸汽发生器的出口位于上端，所述蒸汽发生器的入口位于下端。
  8. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述蒸汽 发生器与所述安全壳内的反应堆压力容器相连接，且所述蒸汽发生器还分别连接主给水管线及主蒸汽管线。
  9. 如权利要求8所述的二次侧非能动余热导出系统，其特征在于：所述主给水管线上设有第四阀门，所述主蒸汽管线上设有第五阀门，所述第四阀门、所述第五阀门均位于所述安全壳内。
  10. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述冷却水箱呈敞口设置。
  11. 如权利要求1所述的二次侧非能动余热导出系统，其特征在于：所述蒸汽发生器为一次通过型蒸汽发生器。
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CN107403650B (zh) * 2017-08-25 2023-11-03 中国船舶重工集团公司第七一九研究所 海上浮动核电站的二次侧非能动余热排出系统

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