WO2021138806A1 - 核电厂严重事故的应对安全系统及其控制方法 - Google Patents

核电厂严重事故的应对安全系统及其控制方法 Download PDF

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Publication number
WO2021138806A1
WO2021138806A1 PCT/CN2020/070695 CN2020070695W WO2021138806A1 WO 2021138806 A1 WO2021138806 A1 WO 2021138806A1 CN 2020070695 W CN2020070695 W CN 2020070695W WO 2021138806 A1 WO2021138806 A1 WO 2021138806A1
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WO
WIPO (PCT)
Prior art keywords
water injection
water
reactor
pit
safety
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PCT/CN2020/070695
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English (en)
French (fr)
Inventor
展德奎
赵鑫海
夏少雄
陈鹏
符卉
吴梓杰
Original Assignee
中广核研究院有限公司
中国广核集团有限公司
中国广核电力股份有限公司
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Application filed by 中广核研究院有限公司, 中国广核集团有限公司, 中国广核电力股份有限公司 filed Critical 中广核研究院有限公司
Priority to PCT/CN2020/070695 priority Critical patent/WO2021138806A1/zh
Priority to CN202080028115.5A priority patent/CN113661547B/zh
Priority to GB2200645.6A priority patent/GB2606803A/en
Priority to EP20912993.1A priority patent/EP3985685A4/en
Publication of WO2021138806A1 publication Critical patent/WO2021138806A1/zh

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Classifications

    • 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
    • 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 technical field of nuclear power, in particular to a safety system for responding to serious accidents in a nuclear power plant and a control method thereof.
  • the nuclear power plant's primary circuit system loses water and the core is exposed and eventually melted.
  • the melted core will eventually collapse into the lower head of the RPV (reactor pressure vessel). If the core melt cannot be cooled in time, due to the core decay heat, the melt will eventually melt through the wall of the RPV lower head, causing the core melt to fall into the pit and may melt through the containment floor, eventually leading to A large amount of radioactive material has leaked.
  • RPV reactor pressure vessel
  • Existing pressurized water reactor nuclear power plants are designed with two or more safety injection boxes.
  • the safety injection boxes are divided into gas space (approximately 25m 3 ) and boric acid water (approximately 35m 3 ) space.
  • the pressure in the primary circuit of nuclear power plants is low. At a certain value, it is generally 4.0-5.0MPa that the electric valve is automatically opened, and boric acid water is quickly injected into the reactor pressure vessel (RPV) through the cold pipe section at one time. But in other cases, there will be no water available in the safety injection box, which will lead to accidents of core melting.
  • RSV reactor pressure vessel
  • the technical problem to be solved by the present invention is to provide a safety system for responding to serious accidents of nuclear power plants and a safety control method for responding to serious accidents of nuclear power plants to improve the safety of nuclear reactors.
  • the technical solution adopted by the present invention to solve its technical problems is to provide a safety system for responding to serious accidents in nuclear power plants, including at least one in-reactor water injection system for injecting water into the reactor pressure vessel;
  • the in-stack water injection system includes a multi-stage safety injection tank and a first water injection pipeline connected to the cold pipe section of the primary circuit system;
  • the internal space of the multi-stage safety injection tank includes a gas phase space distributed from top to bottom and a first-stage water injection space And a second-stage water injection space;
  • the multi-stage safety injection box is provided with a first flow line and a second flow line with a pipe diameter smaller than the first flow line, and the first flow line communicates with the first-stage water injection
  • the space is connected to the first water injection line, and the second flow line communicates with the second stage water injection space and the first water injection line.
  • the pipe diameter of the first flow line is ⁇ 100mm; the pipe diameter of the second flow line is 40mm-80mm, and the water injection flow rate is 20m 3 /h -60m 3 /h.
  • the pressure of the gas phase space is 4.0 MPa-5.0 MPa;
  • the corresponding nuclear power plant primary circuit system pressure is 4.0MPa-5.0MPa;
  • the pressure of the primary circuit system of the nuclear power plant corresponding to the start of water injection in the second-stage water injection space is 0.4MPa-1.0MPa.
  • the in-stack water injection system further includes a first check valve provided on the first water injection line, a first power valve provided on the first flow line, and a first power valve provided on the second flow line. On the second power valve.
  • the in-stack water injection system further includes a level gauge provided on the multi-stage safety injection tank corresponding to the first-stage water injection space;
  • the in-stack water injection system further includes a pressure test gauge and a high-pressure gas source connected to the gas phase space; a third power valve is provided on the connecting pipeline between the high-pressure gas source and the gas phase space;
  • a signal to activate the third power valve is triggered, the third power valve is opened, and the high-pressure gas source supplies air to the multi-stage safety injection box.
  • the internal space of the multi-stage safety injection box further includes at least one third-stage water injection space located below the second-stage water injection space; at least one third flow pipeline is also provided on the multi-stage safety injection box
  • the third flow pipeline communicates with the third-stage water injection space and has a pipe diameter smaller than that of the first flow pipeline.
  • the safety system for responding to serious accidents in the nuclear power plant further includes at least one external cooling water injection system for injecting water into the pit;
  • the external cooling water injection system includes a high-level water injection tank arranged higher than the reactor pressure vessel, and a second water injection pipeline connected between the high-level water injection tank and the reactor pit.
  • the off-stack cooling water injection system further includes a third flow line connected between the high-level water injection tank and the second water injection line; the diameter of the third flow line is smaller than that of the second water injection line The connection position of the third flow line on the high-level water injection tank is lower than the connection position of the second water injection line on the high-level water injection tank.
  • the off-stack cooling water injection system further includes a fourth power valve and a second check valve arranged on the second water injection pipeline in order along the direction from the high-level water injection tank to the pit, and The fifth power valve on the third flow pipeline;
  • connection position of the third flow line on the second water injection line is located between the fourth power valve and the second check valve.
  • the safety system for responding to a serious accident in a nuclear power plant further includes a pit water injection circulating cooling system arranged between the reactor pressure vessel and the pit; the reactor pressure vessel is suspended in the pit and the periphery of the reactor pressure vessel is provided with heat preservation Diversion layer
  • the pit water injection circulation cooling system includes a cooling water channel formed between the reactor pressure vessel and the thermal insulation guide layer, a reactor pit water injection space formed between the thermal insulation guide layer and the inner wall surface of the pit, and is set at the reactor pressure
  • the bottom of the heat preservation guide layer is provided with a water inlet connecting the cooling water flow channel and the water injection space of the pit, so that the cooling water in the water injection space of the pile pit enters the cooling water flow path through the water inlet;
  • the upper end is provided with a steam outlet, and the steam outlet communicates with the cooling water flow channel and the annular reservoir.
  • a buoyancy opening member is provided on the water outlet of the return water channel communicating with the water injection space of the pit;
  • the water outlet remains normally closed; under severe accident conditions of the reactor, the buoyancy opening member is automatically opened after the pit is filled with water.
  • the buoyancy opening member is a buoyancy ball or a buoyancy cover plate.
  • the water inlet of the heat-insulating diversion layer is kept normally closed under normal operating conditions of the reactor; under severe accident conditions of the reactor, the water inlet is open.
  • the present invention also provides a safety control method for responding to serious accidents in nuclear power plants, which adopts any of the above-mentioned safety system for responding to serious accidents in nuclear power plants;
  • the safety control method for responding to serious accidents in nuclear power plants includes:
  • the reactor water injection system Before the severe accident condition of the reactor, when the pressure of the reactor primary circuit system is lower than the first set value, the reactor water injection system quickly injects water into the reactor pressure vessel through the first flow line, the first water injection line and the cold pipe section, and the reactor The core is submerged; when the pressure of the primary loop system is lower than the second set value, the in-reactor water injection system injects water into the reactor pressure vessel through the second flow line, the first water injection line and the cold pipe section to submerge the core; The first setting value is higher than the second setting value.
  • the method further includes: when the reactor is in a severe accident condition, the external cooling water injection system injects water into the reactor pit through the second water injection pipeline.
  • the reactor pit is continuously supplemented with water through the third flow line; the supplemental water flow rate is 30m3/h-70m3/h.
  • the cooling water injected into the pit water injection space of the reactor pit enters the cooling water flow channel through the water inlet of the heat insulating guide layer, and is heated outside the reactor pressure vessel to form a two-phase flow of steam and water ,
  • the two-phase flow of steam and water flows upward along the cooling water channel, and after passing through the steam outlet, the steam and water are separated, and the liquid water falls into the annular reservoir; the cooling water in the annular reservoir returns to the water injection space of the pit through the return flow channel to This loop.
  • the installation of the water injection system inside the reactor can cooperate with the outside cooling water injection system to cool inside and outside the reactor, and effectively cope with the risk of melting the core melt through the RPV after a serious accident.
  • the in-reactor water injection system uses a combination of high-flow and small-flow water injection to extend the time of water injection in the reactor, ensure that the core is in a submerged state for a long time, and significantly slow down the reactor core due to loss of water after an accident.
  • the probability of core degradation and melting delays the progress of serious accidents and maintains the integrity of RPV after serious accidents.
  • Fig. 1 is a schematic structural diagram of a safety system for responding to serious accidents in a nuclear power plant according to an embodiment of the present invention
  • FIG 2 is a schematic diagram of the structure of the in-stack water injection system in the safety response system shown in Figure 1;
  • Fig. 3 is a schematic diagram of the structure of the pit water injection circulating cooling system in the safety response system shown in Fig. 1.
  • the safety system for responding to serious accidents in a nuclear power plant according to an embodiment of the present invention is set in the containment.
  • the safety system includes at least one in-reactor water injection system 10 for injecting water into the reactor pressure vessel 1.
  • the in-stack water injection system 10 includes a multi-stage safety injection tank 11 and a first water injection pipeline 12 connected to the cold pipe section 301 of the primary circuit system 3.
  • the internal space of the multi-stage safety injection box 11 includes a gas-phase space 101 distributed from top to bottom, a first-stage water injection space 102 and a second-stage water injection space 103.
  • the gas phase space 101 is used as a nitrogen space, and the pressure is 4.0 MPa-5.0 MPa.
  • the first-stage water injection space 102 and the second-stage water injection space 103 are liquid phase spaces, used to store cooling water (boron-containing water such as boric acid water), as the medium-pressure stage water injection and the low-pressure stage water injection, respectively.
  • the first-stage water injection space 102 starts water injection
  • the corresponding nuclear power plant primary circuit system pressure is 4.0 MPa-5.0 MPa, that is, when the nuclear power plant’s primary circuit system pressure is 4.0 MPa-5.0 MPa
  • the first stage water injection space 102 starts Water injection.
  • the pressure of the primary circuit system of the nuclear power plant when the second-stage water injection space 103 starts water injection is 0.4 MPa-1.0 MPa, that is, when the pressure of the primary circuit system of the nuclear power plant is 0.4 MPa-1.0 MPa, the second-stage water injection space 103 starts. Water injection.
  • the multi-stage safety injection box 11 is provided with a first flow line 111 and a second flow line 112.
  • the first flow line 111 communicates with the first stage water injection space 102 and the first water injection line 12, and the second flow line 112 communicates with the second stage water injection.
  • the diameter of the first flow line 111 is larger than that of the second flow line 112, so that the first flow line 111 is a large flow pipe, which can quickly inject the cooling water in the first-stage water injection space 102 into the reactor
  • the second flow pipeline 112 is a small flow pipeline, so that the primary loop system 3 can inject water into the core in a low pressure state.
  • the pipe diameter of the first flow line 111 is ⁇ 100mm, and the flow rate is ⁇ 300m 3 /h; the pipe diameter of the second flow line 112 is 40mm-80mm, and the water injection flow rate is 20m 3 /h-60m 3 /h.
  • the first flow line 111 may be the same as the first water injection line 12.
  • the first flow line 111 may correspond to the lower end of the first-stage water injection space 102, one end is connected to the multi-stage safety injection box 11 and communicates with the first-stage water injection space 102, and the other end is connected to and communicated with the first water injection line 12.
  • the second flow line 112 may correspond to the second-stage water injection space 103, one end is connected to the lower end or bottom of the multi-stage safety injection box 11, and the other end is connected to and communicated with the first water injection line 12.
  • the in-stack water injection system 10 also includes a first check valve 13 arranged on the first water injection line 12, a first power valve 14 arranged on the first flow line 111, and a first power valve 14 arranged on the second flow line 112.
  • the first check valve 13 is arranged on the first water injection line 12 to prevent the cooling water from flowing back.
  • the first power valve 14 is used to control the on and off of the first flow line 111
  • the second power valve 15 is used to control the on and off of the second flow line 112.
  • the first power valve 14 and the second power valve 15 are electric valves connected to the instrument control system of the nuclear power plant.
  • the instrument control system controls the opening and closing of the first power valve 14 or the second power valve 15 through automatic signals to achieve Cooling water is automatically injected into the reactor.
  • the first power valve 14 or the second power valve 15 can be opened manually to achieve water injection.
  • the in-stack water injection system 10 also includes a level gauge 16 provided on the multi-stage safety injection box 11 corresponding to the first-stage water injection space 102 to monitor the liquid level of the first-stage water injection space 102.
  • a liquid level alarm is triggered, and a signal to close the first power valve 14 is triggered.
  • the first power valve 14 is automatically closed and the first power valve 14 is automatically closed.
  • the communication between the first-stage water injection space 102 and the first flow line 111 can prevent the gas leakage in the multi-stage safety injection tank 11 from causing insufficient back pressure at the start of the second stage, and enter the reactor through the first flow line 11.
  • the position of the preset value of the liquid level is higher than the connection position of the first flow line 111 on the multi-stage safety injection box 11 to ensure that the gas in the multi-stage safety injection box 11 does not leak.
  • the nuclear power plant's instrument and control system activates the first power valve 14 through an automatic signal, and passes through the first flow line 111 and the cold pipe section 301 quickly injects water into the reactor pressure vessel (RPV) to realize the re-submersion of the core under accident conditions;
  • the liquid level in the multi-stage safety injection tank 11 is lower than the preset value, the liquid level alarm is triggered and the first power valve 14 signal is triggered to close , The first power valve 14 is automatically closed; when the pressure of the primary circuit system 3 continues to drop below the second set value (0.4MPa-1MPa), the signal to start the second power valve 15 is triggered, and the second power valve 15 automatically Start up, and inject water into the reactor pressure vessel through the second flow line 112, and implement the re-submersion of the primary circuit system 3 in a low-pressure state.
  • the first set value such as 4.0 MPa -5.0 MPa
  • the internal space of the multi-stage safety injection box 11 may further include at least one third-stage water injection space (not shown) located below the second-stage water injection space 103 as a subsequent stage water injection space.
  • at least one third flow line (not shown) is also provided on the multi-stage safety injection box 11, and the third flow line communicates with the third-stage water injection space and has a pipe diameter smaller than that of the first flow line 111.
  • the pipe diameter and flow rate of the third flow line may be the same as the second flow line 112.
  • the third flow pipeline is also provided with a power valve to control its on and off.
  • the pressure of the primary circuit system of the nuclear power plant corresponding to the start of water injection in the third stage water injection space is less than the pressure of the primary circuit system of the nuclear power plant corresponding to the start of water injection in the second stage water injection space.
  • the in-stack water injection system 10 also includes a pressure test gauge (not shown) connected to the gas phase space 101 and a high-pressure gas source 17 (such as a high-pressure gas tank); the connecting pipeline 171 between the high-pressure gas source 17 and the gas phase space 101 is provided There is a third power valve 18.
  • the third power valve 18 can be an electric valve.
  • the pressure test gauge and the third power valve 18 are both connected to the instrument control system of the nuclear power plant.
  • the instrument control system controls the opening and closing of the third power valve 18 through the pressure signal to inflate the gas phase space 101 Boost.
  • the gas pressure of the multi-stage safety injection box 11 is less than 2.0 MPa, a signal to activate the third power valve 18 is triggered, the third power valve 18 is opened, and the high-pressure gas source 17 supplies air to the multi-stage safety injection box 11.
  • in-stack water injection systems 10 There may be two or more in-stack water injection systems 10, which are connected to each cold pipe section 301 respectively.
  • the external cooling water injection system 20 includes a high-level water injection tank 21 arranged higher than the reactor pressure vessel 1, and a second water injection pipeline 22 connected between the high-level water injection tank 21 and the pit 2.
  • the high-position water injection tank 21 is used to store cooling water (boron-containing water), and can be injected into the pit 2 by gravity without a power pump.
  • the second water injection pipeline 22 is a pipe with a pipe diameter ⁇ 100 mm, which can quickly fill the pile pit 2 with cooling water.
  • the off-stack cooling water injection system 20 further includes a third flow line 23 connected between the high-level water injection tank 21 and the second water injection line 22.
  • the diameter of the third flow line 23 is smaller than that of the second water injection line 22, and the connection position of the third flow line 23 on the high-level water injection tank 21 is lower than the connection position of the second water injection line 22 on the high-level water injection tank 21.
  • the second water injection line 22 can be connected to the middle 21 or the lower end of the high-level water injection tank 21, and the cooling water is injected into the pit 2 through the second water injection line 22 until the liquid level drops below the inlet water of the second water injection line 22. It can be stopped at the end.
  • the third flow line 23 is connected to the bottom of the high-level water injection tank 21, and the subsequent cooling water can be used to continuously make up water for the pit 2 through the third flow line 23 with a relatively small flow.
  • the flow of the third flow line 23 can be 30m 3 /h-70m 3 /h.
  • the off-stack cooling water injection system 20 also includes a fourth power valve 24 and a second check valve 26 that are sequentially arranged on the second water injection line 22 along the direction from the high-level water injection tank 21 to the pile pit 2, and are arranged at the third flow rate.
  • the fifth power valve 25 on the pipeline 23; the connection position of the third flow pipeline 23 on the second water injection pipeline 22 is located between the fourth power valve 24 and the second check valve 26.
  • the second check valve 26 is provided on the second water injection line 22 to prevent the cooling water from flowing back.
  • the fourth power valve 24 is used to control the on and off of the end of the second water injection pipeline 22 connected to the high-level water injection tank 21, and the fifth power valve 25 is used to control the on and off of the third flow pipeline 23.
  • the fourth power valve 24 and the fifth power valve 25 adopt electric valves, which are connected to the instrument control system of the nuclear power plant.
  • the instrument control system controls the opening and closing of the fourth power valve 24 or the fifth power valve 25 through automatic signals to achieve Cooling water is automatically injected into the pile pit 2.
  • the fourth power valve 24 and the fifth power valve 25 can be opened manually to achieve water injection.
  • the safety system for responding to serious accidents in a nuclear power plant of the present invention also includes a pit water injection circulating cooling system 30 arranged between the reactor pressure vessel 1 and the pit 2.
  • the reactor pressure vessel 1 is suspended in the pit 2 and the outer periphery of the reactor pressure vessel 1 is provided with a heat-preserving and guiding layer 4; the pit 2 is surrounded by a shielding wall.
  • the pit water injection circulating cooling system 30 includes a cooling water flow channel 31 formed between the reactor pressure vessel 1 and the thermal insulation guide layer 4, and formed on the inner wall surface of the thermal insulation guide layer 4 and the pit 2.
  • the bottom of the heat insulating guide layer 4 is provided with a water inlet 41 connecting the cooling water channel 31 and the pit water injection space 32 so that the cooling water in the pit water injection space 32 enters the cooling water channel 32 through the water inlet 41.
  • the water inlet 41 of the thermal insulation guide layer 4 remains normally closed under normal operating conditions of the reactor; under severe accident conditions of the reactor, the water inlet 41 is opened.
  • An exhaust port (not shown) is provided at the upper end of the heat insulating guide layer 4, and the exhaust port is connected to the cooling water channel 31 and the annular reservoir 32.
  • the steam outlet is always located above the liquid level of the annular reservoir 32, and the distance between the center of the steam outlet and the liquid level can be 0.1-0.8m.
  • the heat-insulating flow guide layer 4 includes a flow guide plate and a heat-preserving layer sequentially arranged outside the reactor pressure vessel 1.
  • the second water injection pipeline 22 of the external cooling water injection system 20 is connected to the annular reservoir 33 or the pile pit water injection space 32, and fills the pile pit 2 with cooling water, and is filled with the cooling water flow channel 31 and the pile pit water injection space 32 And the annular reservoir 33.
  • a buoyancy opening member 341 is provided on the outlet of the backwater channel 34 connected to the pit water injection space 32; it is kept normally closed under normal operating conditions of the reactor to reduce the bypass of the ventilation system; under severe accident conditions of the reactor, the pit is
  • the buoyancy opening member 341 can be automatically opened after being filled with water.
  • the buoyancy opening member 341 may be a buoyancy ball or a buoyancy cover plate. After cooling water is injected into the pit water injection space 32, the buoyancy ball floats up to open the water outlet.
  • the water inlet 41 of the heat insulating guide layer 4 is also provided with passive opening and closing parts, such as a buoyancy cover, which floats up to open the water inlet 41 when there is water, and closes when there is no water.
  • the cooling water is heated outside the reactor pressure vessel 1 to form a two-phase flow of steam and water.
  • the two-phase flow of steam and water flows upward along the cooling water channel 31.
  • the steam and water separate, and the liquid water falls into the ring.
  • a natural circulation is formed by using the density difference between the water in the pit 2 and the return flow channel 34, and the natural circulation flow can reach more than 3000m 3 /h, which improves the cooling capacity of the outer wall of the reactor pressure vessel 1;
  • the two-phase flow of soda and water discharged from the steam exhaust hole is automatically separated above the annular reservoir 33, and the steam is discharged to the large space inside the containment through the pores between the main pipe at the upper end of the reactor pressure vessel 1 and the pit shielding wall .
  • the safety control method for responding to severe accidents in nuclear power plants of the present invention adopts the aforementioned safety system for responding to severe accidents in nuclear power plants.
  • the safety control methods for responding to serious accidents in this nuclear power plant can include:
  • the water injection system 10 in the reactor passes through the first flow line 111 and the first water injection line. 12 and cold pipe section 301 quickly inject water into the reactor pressure vessel 1 to submerge the reactor core; when the pressure of the primary loop system 3 is lower than the second set value (such as 0.4MPa-1.0MPa), the reactor water injection system 10 passes through the second The flow line 112, the first water injection line 12, and the cold pipe section 301 inject water into the reactor pressure vessel 1 to submerge the reactor core.
  • the first setting value is higher than the second setting value.
  • the external cooling water injection system 20 injects water into the reactor pit 2 through the second water injection pipeline 22.
  • the pit 2 is continuously supplied with water through the third flow line 23; the make-up water flow is 30m 3 /h-70m 3 /h.
  • the cooling water injected into the pit water injection space 32 of the pit 3 enters the cooling water flow channel 31 through the water inlet of the insulating guide layer 4, and is heated outside the reactor pressure vessel 1 to form soda water Two-phase flow, the two-phase flow of steam and water flows upwards along the cooling water channel 31, and the steam and water are separated after passing through the steam outlet, and the liquid water falls into the annular reservoir 33; the cooling water in the annular reservoir 33 returns through the return water channel 34
  • the pile pit fills the water space 32 to circulate in this way.
  • the in-reactor water injection system is used in conjunction with the external cooling water injection system to perform internal and external cooling. Before a serious accident occurs, the reactor is cooled through the internal water injection system, and the external cooling water injection system can also be extended. Start time.
  • the water injection start time of the external cooling water injection system can be extended to more than 3.5 hours after the accident, and the corresponding external water injection flow can be between 180m 3 /h-360 m 3 /h, the start-up time and the water injection flow outside the reactor are coupled with each other.
  • the free volume of the pit is 180m 3
  • the filling time is shorter.
  • the current filling time is 30 minutes
  • the corresponding flow rate is 360 m 3 /h
  • the start-up time can be up to 4 after the accident.

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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

一种核电厂严重事故的应对安全系统及其控制方法，应对安全系统包括至少一用于向反应堆压力容器（1）内注水的堆内注水系统（10）；堆内注水系统（10）包括多阶段安注箱（11）、连接一回路系统（3）冷管段（301）的第一注水管线（12）；多阶段安注箱（11）的内部空间包括自上而下分布的气相空间（101）、第一阶段注水空间（102）以及第二阶段注水空间（103）；多阶段安注箱（11）上设有第一流量管线（111）和管径小于第一流量管线（111）的第二流量管线（112），第一流量管线（111）连通第一阶段注水空间（102）和第一注水管线（12），第二流量管线（112）连通第二阶段注水空间（103）和第一注水管线（12）。该核电厂严重事故的应对安全系统以堆内注水系统（10）配合堆外冷却注水系统（20）进行堆内、堆外冷却，实现严重事故后维持反应堆压力容器（1）的完整性。

Description

核电厂严重事故的应对安全系统及其控制方法 技术领域
本发明涉及核电技术领域，尤其涉及一种核电厂严重事故的应对安全系统及其控制方法。
 背景技术
在核电厂发生严重事故以后，由于核电厂一回路系统失水而导致堆芯裸露并最终熔化，堆芯熔融物将最终坍塌到RPV（反应堆压力容器）下封头内。如果堆芯熔融物不能及时得到冷却，由于堆芯衰变热的原因，熔融物将最终熔穿RPV下封头壁面，导致堆芯熔融物坠落在堆坑内并有可能熔穿安全壳底板，最终导致放射性物质大量泄漏。
目前国内、外压水堆核电厂大多采用堆坑注水系统来实现堆坑注水，并在反应堆压力容器外壁面形成强迫或者自然循环冷却带走衰变热，防止下封头被熔穿。但是，在大、中破口为始发事件的严重事工况，由于一回路边界破口面积较大，反应堆压力容器失水速度较快，由于堆芯熔化进程的不确定性以及人员启动时间延缓等因素，单纯采用反应堆压力容器外壁面冷却的方式仍存在堆芯熔融物熔穿RPV的可能性。
现有压水堆核电厂均设计2列或者2列以上安注箱，安注箱分为气空间（约为25m 3）及硼酸水（约为35m 3）空间，在核电厂一回路压力低于某特定值时，一般为4.0-5.0MPa自动开启电动阀门，一次性将硼酸水通过冷管段快速注入至反应堆压力容器（RPV）内。但在其他情况下，安注箱内将无水可用，所以会导致堆芯熔化的事故发生。
技术问题
本发明要解决的技术问题在于，提供一种提高核反应堆安全性的核电厂严重事故的应对安全系统及核电厂严重事故的应对安全控制方法。
 技术解决方案
本发明解决其技术问题所采用的技术方案是：提供一种核电厂严重事故的应对安全系统，包括至少一用于向反应堆压力容器内注水的堆内注水系统；
所述堆内注水系统包括多阶段安注箱、连接一回路系统冷管段的第一注水管线；所述多阶段安注箱的内部空间包括自上而下分布的气相空间、第一阶段注水空间以及第二阶段注水空间；所述多阶段安注箱上设有第一流量管线和管径小于所述第一流量管线的第二流量管线，所述第一流量管线连通所述第一阶段注水空间和第一注水管线，所述第二流量管线连通所述第二阶段注水空间和第一注水管线。
优选地，所述第一流量管线的管径≥100mm；所述第二流量管线的管径为40mm-80mm，注水流量20m 3/h -60m 3/h。
优选地，所述气相空间的压力为4.0MPa-5.0MPa；
所述第一阶段注水空间启动注水时所对应的核电厂一回路系统压力为4.0MPa-5.0MPa；
所述第二阶段注水空间启动注水时所对应的核电厂一回路系统压力为0.4MPa-1.0MPa。
优选地，所述堆内注水系统还包括设置在所述第一注水管线上的第一止回阀、设置在所述第一流量管线上的第一动力阀、设置在所述第二流量管线上的第二动力阀。
优选地，所述堆内注水系统还包括对应所述第一阶段注水空间设置在所述多阶段安注箱上的液位计；
在所述多阶段安注箱内液位下降到预设值时，触发关闭所述第一动力阀的信号，所述第一动力阀关闭，断开所述第一流量管线和第一阶段注水空间的连通；所述预设值所在位置高于所述第一流量管线在所述多阶段安注箱上的连接位置。
优选地，所述堆内注水系统还包括连接所述气相空间的压力测试表和高压气源；所述高压气源与所述气相空间的连接管路上设有第三动力阀；在所述多阶段安注箱的气体压力小于2.0MPa时，触发启动所述第三动力阀的信号，所述第三动力阀开启，所述高压气源向所述多阶段安注箱内补气。
优选地，所述多阶段安注箱的内部空间还包括至少一个位于所述第二阶段注水空间下方的第三阶段注水空间；所述多阶段安注箱上还设有至少一第三流量管线，所述第三流量管线连通所述第三阶段注水空间且管径小于所述第一流量管线的管径。
优选地，所述核电厂严重事故的应对安全系统还包括至少一用于向堆坑内注水的堆外冷却注水系统；
所述堆外冷却注水系统包括设置位置高于反应堆压力容器的高位注水水箱、连接在所述高位注水水箱和堆坑之间的第二注水管线。
优选地，所述堆外冷却注水系统还包括连接在所述高位注水水箱和第二注水管线之间的第三流量管线；所述第三流量管线的管径小于所述第二注水管线的管径，且所述第三流量管线在所述高位注水水箱上的连接位置低于所述第二注水管线在所述高位注水水箱上的连接位置。
优选地，所述堆外冷却注水系统还包括沿着从所述高位注水水箱到堆坑的方向依序设置在所述第二注水管线上的第四动力阀和第二止回阀、设置在所述第三流量管线上的第五动力阀；
所述第三流量管线在所述第二注水管线上的连接位置位于所述第四动力阀和第二止回阀之间。
优选地，所述核电厂严重事故的应对安全系统还包括设置在反应堆压力容器和堆坑之间的堆坑注水循环冷却系统；反应堆压力容器悬空设置在堆坑内且反应堆压力容器的外围设有保温导流层；
所述堆坑注水循环冷却系统包括形成在反应堆压力容器和保温导流层之间的冷却水流道、形成在保温导流层和堆坑的内壁面之间的堆坑注水空间、设置在反应堆压力容器上端外围的环形蓄水池、设置在坑壁内并连通所述环形蓄水池和堆坑注水空间的回水流道；所述第二注水管线与所述环形蓄水池或堆坑注水空间相接连通；
所述保温导流层的底部设有连通所述冷却水流道和堆坑注水空间的入水口，以使堆坑注水空间内的冷却水通过入水口进入冷却水流道；所述保温导流层的上端设有排汽口，所述排汽口连通所述冷却水流道和环形蓄水池。
优选地，所述回水通道的连通所述堆坑注水空间的出水口上设有浮力开启件；
在反应堆正常运行工况下，所述出水口保持常闭；在反应堆严重事故工况下，堆坑被水注满后所述浮力开启件自动打开。
优选地，所述浮力开启件为浮力球或浮力盖板。
优选地，所述保温导流层的入水口在反应堆正常运行工况下保持常闭；在反应堆严重事故工况下，所述入水口打开。
本发明还提供一种核电厂严重事故的应对安全控制方法，采用以上任一项所述的核电厂严重事故的应对安全系统；所述核电厂严重事故的应对安全控制方法包括：
在反应堆严重事故工况前，在反应堆一回路系统的压力低于第一设定值时，堆内注水系统通过第一流量管线、第一注水管线和冷管段向反应堆压力容器快速注水，将堆芯淹没；在所述一回路系统的压力低于第二设定值时，堆内注水系统通过第二流量管线、第一注水管线和冷管段向反应堆压力容器注水，将堆芯再淹没；所述第一设定值高于所述第二设定值。
优选地，还包括：在反应堆严重事故工况时，堆外冷却注水系统通过第二注水管线向堆坑内注水。
优选地，在反应堆严重事故工况时，所述堆外冷却注水系统将堆坑注水至目标液位后，通过第三流量管线为堆坑持续补水；补水流量为30m3/h-70m3/h。
优选地，在反应堆严重事故工况时，注入所述堆坑的堆坑注水空间内的冷却水通过保温导流层的入水口进入冷却水流道，在反应堆压力容器外部被加热形成汽水两相流，汽水两相流沿着所述冷却水流道向上流动，通过排汽口后汽水分离，液态水落入环形蓄水池；环形蓄水池内冷却水通过回水流道回到堆坑注水空间，以此循环。
 有益效果
本发明的有益效果：以堆内注水系统的设置，可以配合堆外冷却注水系统进行堆内、堆外冷却，有效应对严重事故后堆芯熔融物熔穿RPV的风险。其中，堆内注水系统以大流量和小流量注水相结合的方式，延长了堆内注水的时长，保障了堆芯长时间处于淹没状态，显著减缓了发生事故后堆芯因为失水而导致堆芯降级、熔化的概率，延缓严重事故进程，实现严重事故后维持RPV的完整性。
 附图说明
下面将结合附图及实施例对本发明作进一步说明，附图中：
图1是本发明一实施例的核电厂严重事故的应对安全系统的结构示意图；
图2是图1中所示应对安全系统中堆内注水系统的结构示意图；
图3是图1中所示应对安全系统中堆坑注水循环冷却系统的结构示意图。
 本发明的实施方式
为了对本发明的技术特征、目的和效果有更加清楚的理解，现对照附图详细说明本发明的具体实施方式。
如图1所示，本发明一实施例的核电厂严重事故的应对安全系统，设置在安全壳内，该应对安全系统包括至少一用于向反应堆压力容器1内注水的堆内注水系统10、至少一用于向堆坑2内注水的堆外冷却注水系统20。
其中，结合图1、2，堆内注水系统10包括多阶段安注箱11、连接一回路系统3的冷管段301的第一注水管线12。多阶段安注箱11的内部空间包括自上而下分布的气相空间101、第一阶段注水空间102以及第二阶段注水空间103。气相空间101作为氮气空间，压力为4.0MPa-5.0MPa。第一阶段注水空间102和第二阶段注水空间103为液相空间，用于存储冷却水（含硼水如硼酸水），分别作为中压阶段注水和低压阶段注水。第一阶段注水空间102启动注水时所对应的核电厂一回路系统压力为4.0MPa-5.0MPa，即当核电厂一回路系统的压力为4.0MPa-5.0MPa时，第一阶段注水空间102开始进行注水。第二阶段注水空间103启动注水时所对应的核电厂一回路系统压力为0.4MPa-1.0MPa，即当核电厂一回路系统的压力为0.4MPa-1.0MPa时，第二阶段注水空间103开始进行注水。多阶段安注箱11上设有第一流量管线111和第二流量管线112，第一流量管线111连通第一阶段注水空间102和第一注水管线12，第二流量管线112连通第二阶段注水空间10和第一注水管线12。相比较而言，第一流量管线111的管径大于第二流量管线112的管径，使得第一流量管线111为大流量管道，可以快速将第一阶段注水空间102内的冷却水注入到反应堆的堆芯，第二流量管线112为小流量管道，实现一回路系统3在低压状态下对堆芯注水。
作为选择，第一流量管线111的管径≥100mm，流量≥300m 3/h；第二流量管线112的管径为40mm-80mm，注水流量20m 3/h -60m 3/h。第一流量管线111可与第一注水管线12相同。
第一流量管线111可以对应第一阶段注水空间102的下端，一端连接在多阶段安注箱11上并与第一阶段注水空间102连通，另一端连接第一注水管线12并与其连通。第二流量管线112可以对应第二阶段注水空间103，一端连接在多阶段安注箱11的下端或底部上，另一端连接第一注水管线12并与其连通。
进一步地，堆内注水系统10还包括设置在第一注水管线12上的第一止回阀13、设置在第一流量管线111上的第一动力阀14、设置在第二流量管线112上的第二动力阀15。第一止回阀13在第一注水管线12上的设置，用于防止冷却水回流。第一动力阀14用于控制第一流量管线111的通断，第二动力阀15用于控制第二流量管线112 的通断。
作为选择，第一动力阀14和第二动力阀15采用电动阀，与核电厂的仪控系统连接，仪控系统通过自动信号控制第一动力阀14或第二动力阀15的启闭，实现冷却水自动注入反应堆。当核电厂发生断电等情况下，可以通过人工将第一动力阀14或第二动力阀15打开，实现注水。
堆内注水系统10还包括对应第一阶段注水空间102设置在多阶段安注箱11上的液位计16，用于监测第一阶段注水空间102的液位。当多阶段安注箱11内第一阶段注水空间102的液位下降到预设值时，触发液位报警，触发关闭第一动力阀14的信号，第一动力阀14自动关闭，断开第一阶段注水空间102和第一流量管线111的连通，可以防止多阶段安注箱11内的气体泄漏而导致在第二阶段启动时背压不足，通过第一流量管线11进入反应堆。液位的预设值所在位置高于第一流量管线111在多阶段安注箱11上的连接位置，确保多阶段安注箱11内气体不泄露。
当反应堆一回路系统3压力低于第一设定值（如4.0 MPa -5.0MPa）时，核电厂的仪控系统通过自动信号，启动第一动力阀14，通过第一流量管线111和冷管段301向反应堆压力容器（RPV）快速注水，实现事故条件下堆芯再淹没；当多阶段安注箱11内的液位低于预设值，触发液位报警，触发关闭第一动力阀14信号，第一动力阀14自动关闭；当一回路系统3的压力继续下降至低于第二设定值（0.4MPa-1MPa）时，触发启动第二动力阀15的信号，第二动力阀15自动启动，通过第二流量管线112向反应堆压力容器注水，实施一回路系统3在低压状态下的再淹没。
根据需要，多阶段安注箱11的内部空间还可包括至少一个位于第二阶段注水空间103下方的第三阶段注水空间（未图示），作为后续阶段注水空间。对应地，多阶段安注箱11上还设有至少一第三流量管线（未图示），第三流量管线连通第三阶段注水空间且管径小于第一流量管线111的管径。第三流量管线的管径与流量可与第二流量管线112相同。同理，第三流量管线上也设有动力阀控制其通断。第三阶段注水空间启动注水时所对应的核电厂一回路系统压力小于第二阶段注水空间启动注水时所对应的核电厂一回路系统压力。
进一步地，堆内注水系统10还包括连接气相空间101的压力测试表（未图示）和高压气源17（如高压气罐）；高压气源17与气相空间101的连接管路171上设有第三动力阀18。第三动力阀18可采用电动阀，压力测试表和第三动力阀18均与核电厂的仪控系统连接，仪控系统通过压力信号控制第三动力阀18的启闭，为气相空间101充气升压。在多阶段安注箱11的气体压力小于2.0MPa时，触发启动第三动力阀18的信号，第三动力阀18开启，高压气源17向多阶段安注箱11内补气。
堆内注水系统10可以是两个或以上，分别与每一冷管段301连接。
如图2所示，堆外冷却注水系统20包括设置位置高于反应堆压力容器1的高位注水水箱21、连接在高位注水水箱21和堆坑2之间的第二注水管线22。高位注水水箱21用于存储冷却水（含硼水），可以通过重力作用往堆坑2注水，不需动力泵。
第二注水管线22为管径≥100mm的管道，可以快速将堆坑2注满冷却水。
本发明中，堆外冷却注水系统20还包括连接在高位注水水箱21和第二注水管线22之间的第三流量管线23。
第三流量管线23的管径小于第二注水管线22的管径，且第三流量管线23在高位注水水箱21上的连接位置低于第二注水管线22在高位注水水箱21上的连接位置。例如，第二注水管线22可以连接在高位注水水箱21的中部21或下端位置上，冷却水通过第二注水管线22注入堆坑2，直至液位下降至低于第二注水管线22的进水端时可停止。第三流量管线23连接高位注水水箱21上的底部，后续冷却水可以通过第三流量管线23相对小流量为堆坑2持续补水，第三流量管线23的流量可为30m 3/h-70m 3/h。
堆外冷却注水系统20还包括沿着从高位注水水箱21到堆坑2的方向依序设置在第二注水管线22上的第四动力阀24和第二止回阀26、设置在第三流量管线23上的第五动力阀25；第三流量管线23在第二注水管线22上的连接位置位于第四动力阀24和第二止回阀26之间。第二止回阀26在第二注水管线22上的设置，用于防止冷却水回流。第四动力阀24用于控制第二注水管线22连接高位注水水箱21的一端的通断，第五动力阀25用于控制第三流量管线23 的通断。
作为选择，第四动力阀24和第五动力阀25采用电动阀，与核电厂的仪控系统连接，仪控系统通过自动信号控制第四动力阀24或第五动力阀25的启闭，实现冷却水自动注入堆坑2。当核电厂发生断电等情况下，可以通过人工将第四动力阀24和第五动力阀25打开，实现注水。
进一步地，本发明的核电厂严重事故的应对安全系统还包括设置在反应堆压力容器1和堆坑2之间的堆坑注水循环冷却系统30。反应堆压力容器1悬空设置在堆坑2内且反应堆压力容器1的外围设有保温导流层4；堆坑2由屏蔽墙围接形成。
如图1、3所示，堆坑注水循环冷却系统30包括形成在反应堆压力容器1和保温导流层4之间的冷却水流道31、形成在保温导流层4和堆坑2的内壁面之间的堆坑注水空间32、设置在反应堆压力容器1上端外围的环形蓄水池33、设置在坑壁（屏蔽墙）内并连通环形蓄水池33和堆坑注水空间32的回水流道34。
其中，保温导流层4的底部设有连通冷却水流道31和堆坑注水空间32的入水口41，以使堆坑注水空间32内的冷却水通过入水口41进入冷却水流道32。保温导流层4的入水口41在反应堆正常运行工况下保持常闭；在反应堆严重事故工况下，入水口41打开。保温导流层4的上端设有排汽口（未图示），排汽口连通冷却水流道31和环形蓄水池32。排汽口始终位于环形蓄水池32的液面上方，排汽口中心距与液面之间可相距0.1-0.8m。
具体地，保温导流层4包括依次设置在反应堆压力容器1外的导流板和保温层。
堆外冷却注水系统20的第二注水管线22与环形蓄水池33或堆坑注水空间32相接连通，将冷却水注满到堆坑2内，充满冷却水流道31、堆坑注水空间32和环形蓄水池33。
回水通道34的连通堆坑注水空间32的出水口上设有浮力开启件341；在反应堆正常运行工况下保持常闭，减少通风系统旁通；在反应堆严重事故工况下，堆坑被水注满后能够浮力开启件341自动打开。浮力开启件341可以是浮力球或浮力盖板，在堆坑注水空间32内注入冷却水后，浮力球上浮打开出水口。保温导流层4的入水口41也设置了非能动开合件，如浮力盖板等，在有水的情况下上浮打开入水口41，无水情况下闭合。
在反应堆严重事故工况下，冷却水在反应堆压力容器1外部被加热形成汽水两相流，汽水两相流沿着冷却水流道31向上流动，通过排汽口后汽水分离，液态水落入环形蓄水池33；环形蓄水池33内的冷却水再通过回水流道34流回堆坑注水空间32，以此形成一个自然循环回路。回水流道34设置多个，沿反应堆压力容器1的周向间隔布置在堆坑2的的坑壁（屏蔽墙）内，不占用堆坑空间，同时不会受到保温导流层4变形或泄露或安装间隙对建立堆坑自然循环的影响。
堆坑注水循环冷却系统30中，利用堆坑2和回水流道34中水的密度差形成自然循环，自然循环流量可以达到3000m 3/h以上，提升了反应堆压力容器1外壁面的冷却能力；同时实现从排汽孔排出的汽水两相流在环形蓄水池33上方实现汽水自动分离，蒸汽通过反应堆压力容器1上端的主管道和堆坑屏蔽墙之间的孔隙排出到安全壳内部大空间。
本发明的核电厂严重事故的应对安全控制方法，采用上述的核电厂严重事故的应对安全系统。结合图1-3，该核电厂严重事故的应对安全控制方法可包括：
在反应堆严重事故工况前，在反应堆的一回路系统3的压力低于第一设定值（如4.0MPa-5.0MPa）时，堆内注水系统10通过第一流量管线111、第一注水管线12和冷管段301向反应堆压力容器1快速注水，将堆芯淹没；在一回路系统3的压力低于第二设定值（如0.4MPa-1.0MPa）时，堆内注水系统10通过第二流量管线112、第一注水管线12和冷管段301向反应堆压力容器1注水，将堆芯再淹没。第一设定值高于第二设定值。
在反应堆严重事故工况时，堆外冷却注水系统20通过第二注水管线22向堆坑2内注水。
其中，在反应堆严重事故工况时，堆外冷却注水系统20将堆坑2注水至目标液位后，通过第三流量管线23为堆坑2持续补水；补水流量为30m 3/h-70m 3/h。
具体地，在反应堆严重事故工况时，注入堆坑3的堆坑注水空间32内的冷却水通过保温导流层4的入水口进入冷却水流道31，在反应堆压力容器1外部被加热形成汽水两相流，汽水两相流沿着冷却水流道31向上流动，通过排汽口后汽水分离，液态水落入环形蓄水池33；环形蓄水池33内冷却水通过回水流道34回到堆坑注水空间32，以此循环。
本发明中，通过堆内注水系统配合堆外冷却注水系统进行堆内、堆外冷却，在反应堆严重事故发生前，先通过堆内注水系统对反应堆进行冷却，同时也可以延长堆外冷却注水系统的启动时间。
例如，若多阶段安注箱11的容积增大60m 3，堆外冷却注水系统的注水启动时间可以延长至事故发生后3.5小时以上，相应的堆外注水流量可以介于180m 3/h-360 m 3/h，启动时间与堆外注水流量是相互耦合的。对于堆坑自由容积为180m 3，如果堆外注水流量越大，注满的时间越短，目前注满时间为30min，对应的流量为360 m 3/h，则启动的时间可以至事故后4小时；若堆外注水流量为180m 3/h，则注满的时间为1小时，则启动时间为3.5小时。明显地，以上启动时间，较无堆内注水条件下20min-30min，大大延长，为现场事故处理人员的判断、操作提供非常宽裕的时间。
以上所述仅为本发明的实施例，并非因此限制本发明的专利范围，凡是利用本发明说明书及附图内容所作的等效结构或等效流程变换，或直接或间接运用在其他相关的技术领域，均同理包括在本发明的专利保护范围内。

Claims (18)

  1. 一种核电厂严重事故的应对安全系统，其特征在于，包括至少一用于向反应堆压力容器内注水的堆内注水系统；
    所述堆内注水系统包括多阶段安注箱、连接一回路系统冷管段的第一注水管线；所述多阶段安注箱的内部空间包括自上而下分布的气相空间、第一阶段注水空间以及第二阶段注水空间；所述多阶段安注箱上设有第一流量管线和管径小于所述第一流量管线的第二流量管线，所述第一流量管线连通所述第一阶段注水空间和第一注水管线，所述第二流量管线连通所述第二阶段注水空间和第一注水管线。
  2. 根据权利要求1所述的核电厂严重事故的应对安全系统，其特征在于，所述第一流量管线的管径≥100mm；所述第二流量管线的管径为40mm-80mm，注水流量20m 3/h -60m 3/h。
  3. 根据权利要求1所述的核电厂严重事故的应对安全系统，其特征在于，所述气相空间的压力为4.0MPa-5.0MPa；
    所述第一阶段注水空间启动注水时所对应的核电厂一回路系统压力为4.0MPa-5.0MPa；
    所述第二阶段注水空间启动注水时所对应的核电厂一回路系统压力为0.4MPa-1.0MPa。
  4. 根据权利要求1所述的核电厂严重事故的应对安全系统，其特征在于，所述堆内注水系统还包括设置在所述第一注水管线上的第一止回阀、设置在所述第一流量管线上的第一动力阀、设置在所述第二流量管线上的第二动力阀。
  5. 根据权利要求4所述的核电厂严重事故的应对安全系统，其特征在于，所述堆内注水系统还包括对应所述第一阶段注水空间设置在所述多阶段安注箱上的液位计；
    在所述多阶段安注箱内液位下降到预设值时，触发关闭所述第一动力阀的信号，所述第一动力阀关闭，断开所述第一流量管线和第一阶段注水空间的连通；所述预设值所在位置高于所述第一流量管线在所述多阶段安注箱上的连接位置。
  6. 根据权利要求1所述的核电厂严重事故的应对安全系统，其特征在于，所述堆内注水系统还包括连接所述气相空间的压力测试表和高压气源；所述高压气源与所述气相空间的连接管路上设有第三动力阀；在所述多阶段安注箱的气体压力小于2.0MPa时，触发启动所述第三动力阀的信号，所述第三动力阀开启，所述高压气源向所述多阶段安注箱内补气。
  7. 根据权利要求1所述的核电厂严重事故的应对安全系统，其特征在于，所述多阶段安注箱的内部空间还包括至少一个位于所述第二阶段注水空间下方的第三阶段注水空间；所述多阶段安注箱上还设有至少一第三流量管线，所述第三流量管线连通所述第三阶段注水空间且管径小于所述第一流量管线的管径。
  8. 根据权利要求1-7任一项所述的核电厂严重事故的应对安全系统，其特征在于，所述核电厂严重事故的应对安全系统还包括至少一用于向堆坑内注水的堆外冷却注水系统；
    所述堆外冷却注水系统包括设置位置高于反应堆压力容器的高位注水水箱、连接在所述高位注水水箱和堆坑之间的第二注水管线。
  9. 根据权利要求8所述的核电厂严重事故的应对安全系统，其特征在于，所述堆外冷却注水系统还包括连接在所述高位注水水箱和第二注水管线之间的第三流量管线；所述第三流量管线的管径小于所述第二注水管线的管径，且所述第三流量管线在所述高位注水水箱上的连接位置低于所述第二注水管线在所述高位注水水箱上的连接位置。
  10. 根据权利要求9所述的核电厂严重事故的应对安全系统，其特征在于，所述堆外冷却注水系统还包括沿着从所述高位注水水箱到堆坑的方向依序设置在所述第二注水管线上的第四动力阀和第二止回阀、设置在所述第三流量管线上的第五动力阀；
    所述第三流量管线在所述第二注水管线上的连接位置位于所述第四动力阀和第二止回阀之间。
  11. 根据权利要求8所述的核电厂严重事故的应对安全系统，其特征在于，所述核电厂严重事故的应对安全系统还包括设置在反应堆压力容器和堆坑之间的堆坑注水循环冷却系统；反应堆压力容器悬空设置在堆坑内且反应堆压力容器的外围设有保温导流层；
    所述堆坑注水循环冷却系统包括形成在反应堆压力容器和保温导流层之间的冷却水流道、形成在保温导流层和堆坑的内壁面之间的堆坑注水空间、设置在反应堆压力容器上端外围的环形蓄水池、设置在坑壁内并连通所述环形蓄水池和堆坑注水空间的回水流道；所述第二注水管线与所述环形蓄水池或堆坑注水空间相接连通；
    所述保温导流层的底部设有连通所述冷却水流道和堆坑注水空间的入水口，以使堆坑注水空间内的冷却水通过入水口进入冷却水流道；所述保温导流层的上端设有排汽口，所述排汽口连通所述冷却水流道和环形蓄水池。
  12. 根据权利要求11所述的核电厂严重事故的应对安全系统，其特征在于，所述回水通道的连通所述堆坑注水空间的出水口上设有浮力开启件；
    在反应堆正常运行工况下，所述出水口保持常闭；在反应堆严重事故工况下，堆坑被水注满后所述浮力开启件自动打开。
  13. 根据权利要求12所述的核电厂严重事故的应对安全系统，其特征在于，所述非能动开合件为浮力球或浮力盖板。
  14. 根据权利要求11所述的核电厂严重事故的应对安全系统，其特征在于，所述保温导流层的入水口在反应堆正常运行工况下保持常闭；在反应堆严重事故工况下，所述入水口打开。
  15. 一种核电厂严重事故的应对安全控制方法，其特征在于，采用权利要求1-14任一项所述的核电厂严重事故的应对安全系统；所述核电厂严重事故的应对安全控制方法包括：
    在反应堆严重事故工况前，在反应堆一回路系统的压力低于第一设定值时，堆内注水系统通过第一流量管线、第一注水管线和冷管段向反应堆压力容器快速注水，将堆芯淹没；在所述一回路系统的压力低于第二设定值时，堆内注水系统通过第二流量管线、第一注水管线和冷管段向反应堆压力容器注水，将堆芯再淹没；所述第一设定值高于所述第二设定值。
  16. 根据权利要求15所述的核电厂严重事故的应对安全方法，其特征在于，还包括：在反应堆严重事故工况时，堆外冷却注水系统通过第二注水管线向堆坑内注水。
  17. 根据权利要求16所述的核电厂严重事故的应对安全方法，其特征在于，在反应堆严重事故工况时，所述堆外冷却注水系统将堆坑注水至目标液位后，通过第三流量管线为堆坑持续补水；补水流量为30m 3/h-70m 3/h。
  18. 根据权利要求16所述的核电厂严重事故的应对安全方法，其特征在于，在反应堆严重事故工况时，注入所述堆坑的堆坑注水空间内的冷却水通过保温导流层的入水口进入冷却水流道，在反应堆压力容器外部被加热形成汽水两相流，汽水两相流沿着所述冷却水流道向上流动，通过排汽口后汽水分离，液态水落入环形蓄水池；环形蓄水池内冷却水通过回水流道回到堆坑注水空间，以此循环。
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Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN113972016A (zh) * 2021-10-26 2022-01-25 中国核动力研究设计院 核电厂安全壳外失水事故应对方法、装置、设备及介质

Citations (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US5519743A (en) * 1994-09-02 1996-05-21 Westinghouse Electric Corporation Primary coolant system of a nuclear power plant for providing coolant to a primary loop
CN102867549A (zh) * 2012-09-27 2013-01-09 中国核电工程有限公司 一种能动与非能动相结合的堆腔注水冷却系统
CN103578581A (zh) * 2012-08-03 2014-02-12 韩国原子力硏究院 通过分离的氮气箱加压的安全注入箱系统
CN103971764A (zh) * 2013-02-01 2014-08-06 韩国原子力研究院 多段式安全注入装置和具有该多段式安全注入装置的无源安全注入系统
CN105280249A (zh) * 2015-09-16 2016-01-27 中广核工程有限公司 核电站反应堆压力容器与屏蔽墙的组合结构
CN109243636A (zh) * 2018-11-09 2019-01-18 中广核工程有限公司 核电厂非能动堆腔注水系统

Family Cites Families (9)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US5180543A (en) * 1989-06-26 1993-01-19 Westinghouse Electric Corp. Passive safety injection system using borated water
DE69009367T2 (de) * 1989-09-19 1994-12-01 Mitsubishi Heavy Ind Ltd System zum Abführen der Reaktorkernzerfallswärme in einem Druckwasserreaktor.
KR101447029B1 (ko) * 2013-02-14 2014-10-08 한국원자력연구원 다단 안전주입 장치 및 이를 구비하는 피동안전주입계통
CN104112482B (zh) * 2014-07-30 2016-09-28 中广核研究院有限公司 非能动自流量控制注水系统
WO2017045159A1 (zh) * 2015-09-16 2017-03-23 中广核工程有限公司 核电站反应堆压力容器与屏蔽墙的组合结构
CN105405479B (zh) * 2015-11-06 2017-11-17 中广核工程有限公司 压水堆核电厂综合非能动安全系统
CN105845187A (zh) * 2016-05-18 2016-08-10 中广核研究院有限公司 核电站严重事故缓解系统
KR101785460B1 (ko) * 2016-11-01 2017-10-16 한국원자력연구원 안전주입계통 및 이를 구비하는 원전
CN209045171U (zh) * 2018-09-03 2019-06-28 中广核研究院有限公司 基于atf燃料的反应堆

Patent Citations (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US5519743A (en) * 1994-09-02 1996-05-21 Westinghouse Electric Corporation Primary coolant system of a nuclear power plant for providing coolant to a primary loop
CN103578581A (zh) * 2012-08-03 2014-02-12 韩国原子力硏究院 通过分离的氮气箱加压的安全注入箱系统
CN102867549A (zh) * 2012-09-27 2013-01-09 中国核电工程有限公司 一种能动与非能动相结合的堆腔注水冷却系统
CN103971764A (zh) * 2013-02-01 2014-08-06 韩国原子力研究院 多段式安全注入装置和具有该多段式安全注入装置的无源安全注入系统
CN105280249A (zh) * 2015-09-16 2016-01-27 中广核工程有限公司 核电站反应堆压力容器与屏蔽墙的组合结构
CN109243636A (zh) * 2018-11-09 2019-01-18 中广核工程有限公司 核电厂非能动堆腔注水系统

Cited By (2)

* Cited by examiner, † Cited by third party
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CN113972016A (zh) * 2021-10-26 2022-01-25 中国核动力研究设计院 核电厂安全壳外失水事故应对方法、装置、设备及介质
CN113972016B (zh) * 2021-10-26 2024-01-26 中国核动力研究设计院 核电厂安全壳外失水事故应对方法、装置、设备及介质

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