WO2022047622A1 - 一种核电站熔融物堆内滞留系统 - Google Patents

一种核电站熔融物堆内滞留系统 Download PDF

Info

Publication number
WO2022047622A1
WO2022047622A1 PCT/CN2020/112845 CN2020112845W WO2022047622A1 WO 2022047622 A1 WO2022047622 A1 WO 2022047622A1 CN 2020112845 W CN2020112845 W CN 2020112845W WO 2022047622 A1 WO2022047622 A1 WO 2022047622A1
Authority
WO
WIPO (PCT)
Prior art keywords
reactor
water
water tank
electric valve
pipeline
Prior art date
Application number
PCT/CN2020/112845
Other languages
English (en)
French (fr)
Inventor
夏少雄
展德奎
陈鹏
吴梓杰
符卉
郭超
赵鑫海
梁洪林
Original Assignee
中广核研究院有限公司
中国广核集团有限公司
中国广核电力股份有限公司
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by 中广核研究院有限公司, 中国广核集团有限公司, 中国广核电力股份有限公司 filed Critical 中广核研究院有限公司
Priority to PCT/CN2020/112845 priority Critical patent/WO2022047622A1/zh
Priority to CN202080106605.2A priority patent/CN116391238A/zh
Publication of WO2022047622A1 publication Critical patent/WO2022047622A1/zh

Links

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 present invention relates to the field of nuclear power plants, and more particularly, to a nuclear power plant smelt in-stack retention system.
  • Hualong No. 1 adopts the strategy of smelt retention in the reactor to deal with core melting accidents.
  • Hualong No. 1 adopts the pit water injection system to realize the retention of molten material in the heap, and the water injected into the pit cools the wall of the lower head to prevent the lower head from being melted through.
  • the core melting is a late stage phenomenon of a serious accident, and there are many uncertainties. It is difficult to accurately calculate the heat transfer of the molten pool of the lower head with the existing software and analysis capabilities.
  • the safety margin of the smelt in-reactor retention system is small, and the uncertainty of software analysis has a great impact on the evaluation of the effectiveness of the smelt in-reactor retention strategy, which affects the safety of the reactor. Therefore, it is necessary to improve the existing smelt retention system in order to improve the safety margin of the reactor.
  • the technical problem to be solved by the present invention is, in view of the above-mentioned defects of the prior art, to provide a nuclear power plant smelt in-stack retention system.
  • the technical solution adopted by the present invention to solve the technical problem is as follows: constructing an in-reactor retention system for smelt in a nuclear power plant, including an external water injection system and an in-reactor water injection system.
  • the in-reactor water injection system is used to inject water into the pressure vessel of the reactor during the severe accident stage of the reactor, and the water used by the in-reactor water injection system is boron-containing water.
  • the in-reactor water injection system includes a high-pressure water tank for storing inert gas and boron-containing water, and the water outlet of the high-pressure water tank is connected to the first stop through a pipeline.
  • the first end of the return valve V9, the second end of the first check valve V9 is connected to the inside of the reactor through a pipeline;
  • the pressure of the inert gas in the high-pressure water tank is higher than the standard atmospheric pressure.
  • the high-pressure water tank starts to inject boron-containing water into the reactor. , and inject the inert gas into the high-pressure water tank after the boron-containing water is injected.
  • an electric valve V8 is provided on the pipeline between the water outlet of the high-pressure water tank and the first check valve V9;
  • the electric valve V8 is in a normally open state, and the electric valve V8 is closed during reactor maintenance.
  • the pressure range of the inert gas in the high-pressure water tank is 0.1 MPa to 2 MPa.
  • the pipeline between the water outlet of the high-pressure water tank and the first end of the first check valve V9 and the connection between the first check valve V9 is in the range of 50mm to 80mm.
  • the lowest point of the high-pressure water tank is higher than the highest point of the reactor.
  • the nuclear power plant smelt in-stack retention system of the present invention further comprises a refueling water tank connected to the water inlet pipe of the high-pressure water tank, and the refueling water tank stores boron-containing water;
  • An electric pump P1 and an electric valve V11 are arranged on the pipeline between the water inlet of the high-pressure water tank and the refueling water tank, and the refueling water tank is used to supplement boron-containing water into the high-pressure water tank.
  • the out-of-reactor water injection system includes a high-level atmospheric pressure water tank, an electric valve V4 and a second check valve V6, and the lowest point of the high-level atmospheric pressure water tank is high at the highest point of the reactor;
  • the bottom water outlet of the high-level atmospheric water tank is connected to the first end of the electric valve V4 through a pipeline, and the second end of the electric valve V4 is connected to the first end of the second check valve V6 through a pipeline.
  • the second end of the second check valve V6 is connected to the outside of the reactor through a pipeline;
  • the electric valve V4 is in a closed state when the reactor is working normally, and is opened when the reactor core outlet temperature exceeds 650° C. After opening, the water in the high-level atmospheric water tank is injected into the outside of the reactor.
  • the nuclear power plant smelt retention system in the present invention further includes a backup electric valve V5, the first end of the backup electric valve V5 is connected to the first end of the electric valve V4 through a pipeline, and the backup electric valve V5 The second end of the motor is connected to the second end of the electric valve V4 through a pipeline;
  • the backup electric valve V5 is in a closed state when the reactor is in normal operation.
  • the electric valve V4 is first opened. If the electric valve V4 fails to open, the backup electric valve V5 is opened.
  • the water in the high-level atmospheric water tank is injected outside the reactor.
  • the nuclear power plant smelt retention system in the present invention further includes a shut-off valve V7 arranged on the pipeline between the second end of the second check valve V6 and the outside of the reactor.
  • the shut-off valve V7 is used for To terminate the water injection in the event of mis-injection.
  • the high-level atmospheric water tank includes at least two bottom water outlets, and each bottom water outlet is connected to the first end of the electric valve V4 through a pipeline, An electric valve is correspondingly arranged on the pipeline of each bottom water outlet, and all the electric valves are in a closed state when the reactor is in normal operation, and open when the reactor core outlet temperature exceeds 650°C.
  • the nozzle of each of the bottom water outlets corresponding to the pipeline extends to the interior of the high-level atmospheric pressure water tank, and each of the bottom water outlets corresponds to the pipe opening of the pipeline.
  • the nozzles have different extension heights, and the amount of water above the nozzle of the pipe with the highest extension is equal to the free volume of the reactor pit.
  • the high-level atmospheric pressure water tank includes three bottom water outlets, and the pipeline between the first bottom water outlet and the first end of the electric valve V4 is provided There is an electric valve V1, an electric valve V2 is arranged on the pipeline between the second bottom water outlet and the first end of the electric valve V4, and an electric valve V2 is arranged on the pipeline between the third bottom water outlet and the first end of the electric valve V4 There is an electric valve V3, the electric valve V1, the electric valve V2 and the electric valve V3 are in a closed state when the reactor is working normally, and open when the reactor core outlet temperature exceeds 650 °C;
  • the corresponding pipeline of the first bottom water outlet has no extension in the high-level atmospheric pressure water tank, and the extension height of the corresponding pipeline of the third bottom water outlet in the high-level atmospheric pressure water tank is higher than that of the corresponding pipeline of the second bottom water outlet
  • the extension height in the high-level atmospheric water tank, and the amount of water above the nozzle of the corresponding pipeline of the third bottom water outlet is equal to the free volume of the reactor pit.
  • the nuclear power plant smelt in-stack retention system further includes a refueling water tank connected to the water inlet pipeline of the high-level atmospheric pressure water tank, and between the water inlet of the high-level atmospheric pressure water tank and the refueling water tank An electric pump P1 and an electric valve V10 are arranged on the pipeline of , and the refueling water tank is used to supplement water to the high-level atmospheric pressure water tank.
  • the water in the high-level atmospheric water tank is boron-containing water
  • the water in the refueling water tank is boron-containing water.
  • all electric valves in the system are powered by an uninterruptible power supply.
  • the implementation of a nuclear power plant smelt retention system in the reactor has the following beneficial effects: the invention adopts the simultaneous injection of water inside and outside the reactor to realize the smelt retention in the reactor, and greatly improves the safety margin of the nuclear power plant reactor.
  • FIG. 1 is a schematic structural diagram of a smelt retention system in a nuclear power plant according to an embodiment.
  • the in-reactor retention system for smelt in the nuclear power plant of the present embodiment includes an external-reactor water injection system and an in-reactor water-injection system.
  • the water is injected into the reactor, and the pressure vessel is soaked in water; the in-reactor water injection system is used to inject water into the reactor of the reactor 10 during the serious accident stage of the reactor 10.
  • the water used is boron-containing water.
  • the abnormality of the reactor 10 means that the detection system detects a serious accident signal, and starts water injection after detecting the serious accident signal.
  • the in-reactor water injection system in the in-reactor retention system of the nuclear power plant in this embodiment includes a high-pressure water tank 20 for storing inert gas and boron-containing water, and the water outlet of the high-pressure water tank 20 is connected to the first end of the first check valve V9 through a pipe , the second end of the first check valve V9 is connected to the inside of the reactor 10 through a pipeline.
  • the system is an automatic commissioning system.
  • the pressure of the inert gas in the high-pressure water tank 20 is higher than the standard atmospheric pressure. When the reactor system is in normal operation, the pressure in the pressure vessel is higher than the pressure in the pressurized water tank.
  • the coolant in the pressure vessel is No backflow to pressurized water tank.
  • the primary circuit of the reactor is depressurized, resulting in a gradual decrease in the internal pressure of the reactor 10.
  • the high-pressure water tank 20 begins to pressurize.
  • the boron-containing water is injected into the reactor 10 , and the inert gas in the high-pressure water tank 20 is injected into the boron-containing water after the injection of the boron-containing water is completed.
  • inert gas is used to mix with hydrogen to avoid explosion and improve safety.
  • the inert gas may be low-cost nitrogen, although other inert gases may also be used.
  • an electric valve V8 is provided on the pipeline between the water outlet of the high-pressure water tank 20 and the first check valve V9; the electric valve V8 is in a normally open state, and the electric valve V8 shuts down while reactor 10 is being serviced, preventing the system from automatically being put into operation.
  • the pressure of the inert gas in the high-pressure water tank 20 ranges from 0.1 MPa to 2 MPa.
  • the pipeline between the water outlet of the high-pressure water tank 20 and the first end of the first check valve V9 and the second end of the first check valve V9 to the reactor The pipes between the stacks of 10 have pipe inner diameters ranging from 50mm to 80mm.
  • the existing IVR water tank is used to inject water into the pile pit and the pile at the same time, the water volume of the IVR water tank is limited, and the water injection is carried out by using a pipe with a pipe diameter larger than 80 mm, and the sustainable water injection time is short, so the core melts after the water injection is stopped. The process restarted, and the size of the molten pool quickly returned to the level of no water injection.
  • the lowest point of the high-pressure water tank 20 is higher than the highest point of the reactor 10 .
  • the nuclear power plant smelt retention system in this embodiment further includes a refueling water tank 30 connected to the water inlet pipe of the high-pressure water tank 20, and the refueling water tank 30 stores boron-containing water;
  • An electric pump P1 and an electric valve V11 are arranged on the pipeline between the refueling water tanks 30 , and the refueling water tank 30 is used to replenish boron-containing water into the high-pressure water tank 20 .
  • only the electric valve is driven by the battery, and the entire water injection process is passive, and can continue to operate under the condition of the whole plant power failure.
  • the electric valve V11 is normally closed.
  • the electric pump P1 and the electric valve V11 are turned on to inject water into the high-pressure water tank 20.
  • the air inlet of the high-pressure water tank 20 is correspondingly connected to an inert gas storage tank (not shown in the figure), and the inert gas is injected into the high-pressure water tank 20 by using the inert gas storage tank.
  • the out-of-reactor water injection system includes a high-level atmospheric water tank 40 , an electric valve V4 and a second check valve V6 , and the lowest point of the high-level atmospheric pressure water tank 40 is higher than the highest point of the reactor 10 .
  • the water in the high-level atmospheric water tank 40 can enter the reactor pit without power, so as to avoid the failure of water injection due to the lack of power supply after the reactor has an accident.
  • the water in the high-level atmospheric pressure water tank 40 in the smelt reactor retention system of the nuclear power plant of the present embodiment is boron-containing water.
  • the bottom water outlet of the high-level atmospheric water tank 40 is connected to the first end of the electric valve V4 through a pipeline, the second end of the electric valve V4 is connected to the first end of the second check valve V6 through a pipeline, and the second end of the second check valve V6 is connected through a pipeline.
  • the ends are connected to the outside of the reactor 10 by pipes.
  • the electric valve V4 is closed when the reactor 10 is working normally, and is opened when the core outlet temperature of the reactor 10 exceeds 650°C.
  • the nuclear power plant smelt retention system in this embodiment further includes a backup electric valve V5, the first end of the backup electric valve V5 is connected to the first end of the electric valve V4 through a pipeline, and the second end of the backup electric valve V5 is connected through a pipeline. Connect the second end of the electric valve V4.
  • the backup electric valve V5 is closed when the reactor 10 is working normally.
  • the electric valve V4 is first opened. If the electric valve V4 fails to open, the backup electric valve V5 is opened. The water is injected outside the reactor 10.
  • the nuclear power plant smelt retention system in the present embodiment further includes a shut-off valve V7 disposed on the pipeline between the second end of the second check valve V6 and the outside of the reactor 10, and the shut-off valve V7 is used for the occurrence of Terminate water injection when water is injected by mistake.
  • the high-level atmospheric pressure water tank 40 in the nuclear power plant smelt in-reactor retention system of the present embodiment includes at least two bottom water outlets, each bottom water outlet is connected to the first end of the electric valve V4 through a pipeline, and each bottom water outlet is connected to the first end of the electric valve V4
  • an electric valve is set on the pipeline of the reactor 10, and all the electric valves are in a closed state when the reactor 10 is working normally, and open when the reactor 10 core outlet temperature exceeds 650 °C.
  • the high-level atmospheric pressure water tank 40 needs to be activated, and the heap pit is quickly filled with water within 20-120 minutes, and the water injection flow rate is required to be large.
  • the pile pit is filled with water, the water in the pile pit will be reduced by evaporation.
  • the nozzle of the pipe corresponding to each bottom water outlet in the nuclear power plant smelt in-reactor retention system in this embodiment extends to the interior of the high-level atmospheric pressure water tank 40, and the extension height of the nozzle corresponding to each bottom water outlet is different, that is, more The nozzles of the pipes corresponding to the bottom water outlets are distributed in height, the nozzles of the pipes corresponding to different bottom water outlets are distributed at different heights, and the amount of water above the nozzle of the pipe with the highest extension height is equal to the free volume of the reactor 10. .
  • the high-level atmospheric pressure water tank 40 includes three bottom water outlets, an electric valve V1 is provided on the pipeline between the first bottom water outlet and the first end of the electric valve V4, and the second bottom water outlet is to the first end of the electric valve V4
  • An electric valve V2 is arranged on the pipeline between the two, and an electric valve V3 is arranged on the pipeline between the third bottom water outlet and the first end of the electric valve V4.
  • the electric valve V1, the electric valve V2 and the electric valve V3 are in the normal operation of the reactor 10. It is in a closed state, and is opened after the core outlet temperature of the reactor 10 exceeds 650°C.
  • the corresponding pipeline of the first bottom water outlet does not extend in the high-level atmospheric pressure water tank 40, and the corresponding pipeline of the third bottom water outlet in the high-level atmospheric pressure water tank 40 has a higher extension height than the corresponding pipeline of the second bottom water outlet in the high-level atmospheric pressure.
  • the extension height in the water tank 40 and the amount of water above the nozzle of the pipe corresponding to the third bottom water outlet is equal to the free volume of the reactor pit of the reactor 10 .
  • the nozzle of the pipeline corresponding to the third bottom water outlet is exposed, and water can only be injected from the first bottom water outlet and the second bottom water outlet at the same time, and the water injection speed is relatively reduced; after a period of time, the amount of water evaporation is further If it is lowered, the nozzle of the pipe corresponding to the second bottom water outlet is exposed, and water can only be injected from the first bottom water outlet, and the water injection speed is further reduced.
  • the nuclear power plant smelt retention system in this embodiment further includes a refueling water tank 30 connected to the water inlet pipe of the high-level atmospheric pressure water tank 40 , and a pipeline between the water inlet of the high-level atmospheric pressure water tank 40 and the refueling water tank 30
  • An electric pump P1 and an electric valve V10 are arranged on the top, and the refueling water tank 30 is used to replenish water to the high-level atmospheric water tank 40 .
  • the water in the refueling tank 30 is boron-containing water.
  • the electric valve V10 is normally closed. After a serious accident occurs for a period of time, such as 12 hours, the electric pump P1 and the electric valve V10 are turned on to inject water into the high-level atmospheric water tank 40 .
  • all the electric valves in the nuclear power plant melt reactor retention system in this embodiment are powered by an uninterruptible power supply, that is, the electric valve V1, the electric valve V2, the electric valve V3, the electric valve V4, the electric valve V5, the electric valve V8, the electric valve Valve V10 and electric valve V11 are powered by uninterruptible power supply.
  • the time when the molten pool of the lower head of the reactor pressure vessel (RPV) of the nuclear power plant reaches the maximum decay heat is delayed by about 8 hours compared with the case of no water injection in the reactor, and the decay Heat is correspondingly reduced by nearly 30%.
  • the heat flux density is greatly reduced, and the peak value is reduced by more than 30%, which greatly increases the IVR safety margin.
  • the calculation results show that after the completion of water injection in the reactor, even if the core melting process is restarted, the total amount of molten material in the reactor does not decrease due to the water injection in the reactor, but because the time process is greatly delayed, the decay heat is also greatly reduced.
  • the heat flux density for heat transfer in the molten pool increases the safety margin from 10% to about 50%. This embodiment can significantly improve the safety performance of the reactor.
  • RAM random access memory
  • ROM read only memory
  • EEPROM electrically programmable ROM
  • EEPly erasable programmable ROM registers
  • hard disk removable disk
  • CD-ROM compact disc-read only memory

Landscapes

  • 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

一种核电站熔融物堆内滞留系统。该系统包括堆外注水系统和堆内注水系统，堆外注水系统用于在反应堆（10）严重事故阶段向反应堆（10）的堆外注水，堆内注水系统用于在反应堆（10）严重事故阶段向反应堆（10）的堆内注水，堆内注水系统使用的水为含硼水。采用堆内堆外同时注水来实现熔融物堆内滞留，大幅提升核电站反应堆（10）的安全裕量。

Description

一种核电站熔融物堆内滞留系统 技术领域
本发明涉及核电站领域，更具体地说，涉及一种核电站熔融物堆内滞留系统。
 背景技术
能够应对和缓解堆芯熔化事故为三代反应堆的典型特征，中国自主研发的第三代反应堆“华龙一号”采用熔融物堆内滞留策略来应对堆芯熔化事故。目前华龙一号采用堆坑注水系统来实现熔融物堆内滞留，注入堆坑内的水冷却下封头壁面，防止下封头被熔穿。但堆芯熔化为严重事故晚期现象，存在较多的不确定性，现有软件和分析能力很难准确计算下封头熔池的传热。再加上熔融物堆内滞留系统安全裕量较小，软件分析不确定性对熔融物堆内滞留策略的有效性评估带来较大影响，影响反应堆的安全。因此需要对现有的熔融物堆内滞留系统进行改进，以提升反应堆的安全裕量。
 技术问题
本发明要解决的技术问题在于，针对现有技术的上述缺陷，提供一种核电站熔融物堆内滞留系统。
 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：构造一种核电站熔融物堆内滞留系统，包括堆外注水系统和堆内注水系统，所述堆外注水系统用于在反应堆严重事故阶段向反应堆的堆坑进行注水，所述堆内注水系统用于在反应堆严重事故阶段向反应堆的压力容器内进行注水，所述堆内注水系统使用的水为含硼水。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述堆内注水系统包括用于存储惰性气体和含硼水的高压水箱，所述高压水箱的出水口通过管道连接第一止回阀V9的第一端，所述第一止回阀V9的第二端通过管道连接至反应堆的堆内；
所述高压水箱内惰性气体的压强大于标准大气压，当反应堆的堆内压强小于所述第一止回阀V9的第一端的压强时，所述高压水箱开始向反应堆的堆内注入含硼水，且在含硼水注入完毕后注入所述高压水箱内的惰性气体。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高压水箱的出水口和所述第一止回阀V9之间管道上设置有电动阀V8；
所述电动阀V8处于常开状态，所述电动阀V8在反应堆维修时关闭。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高压水箱内惰性气体的压强范围为0.1MPa至2MPa。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高压水箱的出水口至所述第一止回阀V9的第一端之间管道和所述第一止回阀V9的第二端至反应堆的堆内之间管道的管道内径范围为50mm至80mm。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高压水箱的最低点高于反应堆的最高点。
进一步，本发明所述的核电站熔融物堆内滞留系统还包括与所述高压水箱的入水口管道连接的换料水箱，所述换料水箱内存储有含硼水；
所述高压水箱的入水口和所述换料水箱之间的管道上设置有电动泵P1和电动阀V11，所述换料水箱用于向所述高压水箱内补充含硼水。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述堆外注水系统包括高位常压水箱、电动阀V4和第二止回阀V6，所述高位常压水箱的最低点高于反应堆的最高点；
所述高位常压水箱的底部出水口通过管道连接所述电动阀V4的第一端，所述电动阀V4的第二端通过管道连接所述第二止回阀V6的第一端，所述第二止回阀V6的第二端通过管道连接至反应堆的堆外；
所述电动阀V4在反应堆正常工作时处于关闭状态，在反应堆堆芯出口温度超过650℃后开启，开启后所述高位常压水箱内的水注入反应堆的堆外。
进一步，本发明所述的核电站熔融物堆内滞留系统还包括备用电动阀V5，所述备用电动阀V5的第一端通过管道连接所述电动阀V4的第一端，所述备用电动阀V5的第二端通过管道连接所述电动阀V4的第二端；
所述备用电动阀V5在反应堆正常工作时处于关闭状态，在反应堆堆芯出口温度超过650℃后首先开启所述电动阀V4，若所述电动阀V4开启失败后开启所述备用电动阀V5，所述高位常压水箱内的水注入反应堆的堆外。
进一步，本发明所述的核电站熔融物堆内滞留系统还包括设置在所述第二止回阀V6的第二端和反应堆的堆外之间管道上的截止阀V7，所述截止阀V7用于在发生误注水时终止注水。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高位常压水箱包括至少两个底部出水口，每个底部出水口通过管道连接至所述电动阀V4的第一端，每个底部出水口的管道上对应设置一个电动阀，所有所述电动阀在反应堆正常工作时处于关闭状态，在反应堆堆芯出口温度超过650℃后开启。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，每个所述底部出水口对应管道的管口延伸至所述高位常压水箱的内部，每个所述底部出水口对应管道的管口的延伸高度不同，且延伸高度最高的管道的管口之上的水量等于反应堆的堆坑自由容积。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高位常压水箱包括3个底部出水口，第一底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V1，第二底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V2，第三底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V3，所述电动阀V1、所述电动阀V2和所述电动阀V3在反应堆正常工作时处于关闭状态，在反应堆堆芯出口温度超过650℃后开启；
所述第一底部出水口的对应管道在高位常压水箱内无延伸，所述第三底部出水口的对应管道在高位常压水箱内的延伸高度高于所述第二底部出水口的对应管道在所述高位常压水箱内的延伸高度，且所述第三底部出水口对应管道的管口之上的水量等于反应堆的堆坑自由容积。
进一步，本发明所述的核电站熔融物堆内滞留系统还包括与所述高位常压水箱的入水口管道连接的换料水箱，所述高位常压水箱的入水口和所述换料水箱之间的管道上设置有电动泵P1和电动阀V10，所述换料水箱用于向所述高位常压水箱内补充水。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，所述高位常压水箱内的水为含硼水；
所述换料水箱内的水为含硼水。
进一步，在本发明所述的核电站熔融物堆内滞留系统中，系统中所有电动阀采用不间断电源供电。
 有益效果
实施本发明的一种核电站熔融物堆内滞留系统，具有以下有益效果：本发明采用堆内堆外同时注水来实现熔融物堆内滞留，大幅提升核电站反应堆的安全裕量。
 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是一实施例提供的一种核电站熔融物堆内滞留系统的结构示意图。
 本发明的最佳实施方式
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。
 实施例
参考图1，本实施例的核电站熔融物堆内滞留系统包括堆外注水系统和堆内注水系统，堆外注水系统用于在反应堆10严重事故阶段向反应堆10的堆外注水，堆外注水指将水注入反应堆堆内里，将压力容器泡在水中；堆内注水系统用于在反应堆10严重事故阶段向反应堆10的堆内注水，堆内注水指将水注入反应堆压力容器内，堆内注水系统使用的水为含硼水。反应堆10异常是指检测系统检测到严重事故信号，检测到严重事故信号后启动注水。
本实施例的核电站熔融物堆内滞留系统中堆内注水系统包括用于存储惰性气体和含硼水的高压水箱20，高压水箱20的出水口通过管道连接第一止回阀V9的第一端，第一止回阀V9的第二端通过管道连接至反应堆10的堆内。该系统为自动投运系统，高压水箱20内惰性气体的压强大于标准大气压，反应堆系统正常运行时压力容器内压强大于加压水箱压强，因为存在第一止回阀V9，压力容器内的冷却剂不会回流至加压水箱。严重事故工况下反应堆的一回路卸压，导致反应堆10的堆内压强逐渐减小，当反应堆10的堆内压强小于第一止回阀V9的第一端的压强时，高压水箱20开始向反应堆10的堆内注入含硼水，且在含硼水注入完毕后注入高压水箱20内的惰性气体。因向堆内注水会产生氢气，为防止二次事故，本实施例使用惰性气体与氢气混合，避免发生爆炸，提高安全性。作为选择，惰性气体可选用低成本的氮气，当然也可以选用其他惰性气体。
作为选择，在本实施例的核电站熔融物堆内滞留系统中，高压水箱20的出水口和第一止回阀V9之间管道上设置有电动阀V8；电动阀V8处于常开状态，电动阀V8在反应堆10维修时关闭，防止系统自动投入运行。
作为选择，在本实施例的核电站熔融物堆内滞留系统中，高压水箱20内惰性气体的压强范围为0.1MPa至2MPa。
作为选择，在本实施例的核电站熔融物堆内滞留系统中，高压水箱20的出水口至第一止回阀V9的第一端之间管道和第一止回阀V9的第二端至反应堆10的堆内之间管道的管道内径范围为50mm至80mm。本实施例中由于采用现有IVR水箱进行堆坑、堆内同时注水，IVR水箱的水容积有限，采用大于80mm管径的管道进行注水，可持续注水时间短，因此在注水停止后堆芯熔化进程重启，熔池大小又迅速回升到无注水的水平。采用小于50mm管径的管道进行注水，能提供较长时间的持续冷却，但是由于注水流量小，冷却效果不明显，对熔池的规模减小作用也很小。因此，从堆内注水对熔池的影响来看，50mm至80mm管径注水是最优选择。在该管径下，延迟0-50分钟的注水，对熔池规模的减小作用最明显。
作为选择，在本实施例的核电站熔融物堆内滞留系统中，高压水箱20的最低点高于反应堆10的最高点。
作为选择，本实施例的核电站熔融物堆内滞留系统还包括与高压水箱20的入水口管道连接的换料水箱30，换料水箱30内存储有含硼水；高压水箱20的入水口和换料水箱30之间的管道上设置有电动泵P1和电动阀V11，换料水箱30用于向高压水箱20内补充含硼水。本实施例的核电站熔融物堆内滞留系统仅需电池驱动电动阀，整个注水过程为非能动，可以在全厂断电工况下继续运行。电动阀V11为常闭状态，在严重事故发生12小时后（根据核电站运行规程，事故发生后6小时移动电源就位），开启电动泵P1和电动阀V11向高压水箱20注水。高压水箱20的进气口对应连接惰性气体存储罐（图中未示出），在使用惰性气体存储罐向高压水箱20内注入惰性气体。
在本实施例的核电站熔融物堆内滞留系统中堆外注水系统包括高位常压水箱40、电动阀V4和第二止回阀V6，高位常压水箱40的最低点高于反应堆10的最高点，则高位常压水箱40的水可不需要动力即可进入堆坑，避免反应堆出现事故后，没有电力供应导致无法注水。作为选择，本实施例的核电站熔融物堆内滞留系统中高位常压水箱40内的水为含硼水。高位常压水箱40的底部出水口通过管道连接电动阀V4的第一端，电动阀V4的第二端通过管道连接第二止回阀V6的第一端，第二止回阀V6的第二端通过管道连接至反应堆10的堆外。电动阀V4在反应堆10正常工作时处于关闭状态，在反应堆10堆芯出口温度超过650℃后开启，开启后高位常压水箱40内的水注入反应堆10的堆外。
作为选择，本实施例的核电站熔融物堆内滞留系统还包括备用电动阀V5，备用电动阀V5的第一端通过管道连接电动阀V4的第一端，备用电动阀V5的第二端通过管道连接电动阀V4的第二端。备用电动阀V5在反应堆10正常工作时处于关闭状态，在反应堆10堆芯出口温度超过650℃后首先开启电动阀V4，若电动阀V4开启失败后开启备用电动阀V5，高位常压水箱40内的水注入反应堆10的堆外。
作为选择，本实施例的核电站熔融物堆内滞留系统还包括设置在第二止回阀V6的第二端和反应堆10的堆外之间管道上的截止阀V7，截止阀V7用于在发生误注水时终止注水。
作为选择，本实施例的核电站熔融物堆内滞留系统中高位常压水箱40包括至少两个底部出水口，每个底部出水口通过管道连接至电动阀V4的第一端，每个底部出水口的管道上对应设置一个电动阀，所有电动阀在反应堆10正常工作时处于关闭状态，在反应堆10堆芯出口温度超过650℃后开启。
作为选择，检测到严重事故信号后，需要启动高位常压水箱40，且在20-120分钟内快速将堆坑注满水，此时对注水流量要求较大。当堆坑注满水后，堆坑内的水会因蒸发作用减小，此时需要不断往堆坑注水，以维持自然循环必须的液位。因为水箱容积有限，为了提高系统缓解严重事故能力，注水流量需要与蒸发量相匹配。所以本实施例的核电站熔融物堆内滞留系统中每个底部出水口对应管道的管口延伸至高位常压水箱40的内部，每个底部出水口对应管道的管口的延伸高度不同，即多个底部出水口对应管道的管口在高度上错落分布，不同底部出水口对应管道的管口分布在不同高度，并且延伸高度最高的管道的管口之上的水量等于反应堆10的堆坑自由容积。刚开始注水时，所有底部出水口同时出水，在20-120分钟内快速将堆坑注满水；注满水后，高位常压水箱40的水位下降到延伸高度最高的管道的管口之下，则出水量相应减小。以此类推，随着高位常压水箱40内水位的下降，能够出水的管口越来越少，出水量逐渐较小，以适配蒸发量的减小，保持堆坑内水的平衡。
例如，高位常压水箱40包括3个底部出水口，第一底部出水口至电动阀V4的第一端之间管道上设置有电动阀V1，第二底部出水口至电动阀V4的第一端之间管道上设置有电动阀V2，第三底部出水口至电动阀V4的第一端之间管道上设置有电动阀V3，电动阀V1、电动阀V2和电动阀V3在反应堆10正常工作时处于关闭状态，在反应堆10堆芯出口温度超过650℃后开启。第一底部出水口的对应管道在高位常压水箱40内无延伸，第三底部出水口的对应管道在高位常压水箱40内的延伸高度高于第二底部出水口的对应管道在高位常压水箱40内的延伸高度，且第三底部出水口对应管道的管口之上的水量等于反应堆10的堆坑自由容积。刚开始注水时，第一底部出水口、第二底部出水口和第三底部出水口同时快速注水，短时间内淹没堆坑。堆坑注满水后，第三底部出水口对应管道的管口裸露，仅能由第一底部出水口和第二底部出水口同时注水，则注水速度相对减小；一段时间后水分蒸发量进一步降低，则第二底部出水口对应管道的管口裸露，仅能由第一底部出水口注水，注水速度进一步减小。
作为选择，本实施例的核电站熔融物堆内滞留系统还包括与高位常压水箱40的入水口管道连接的换料水箱30，高位常压水箱40的入水口和换料水箱30之间的管道上设置有电动泵P1和电动阀V10，换料水箱30用于向高位常压水箱40内补充水。作为选择；换料水箱30内的水为含硼水。电动阀V10为常闭状态，在严重事故发生一段时间后，例如12小时，开启电动泵P1和电动阀V10向高位常压水箱40注水。
作为选择，本实施例的核电站熔融物堆内滞留系统中所有电动阀采用不间断电源供电，即电动阀V1、电动阀V2、电动阀V3、电动阀V4、电动阀V5、电动阀V8、电动阀V10、电动阀V11采用不间断电源供电。
本实施例采用本实施例的堆内堆外同时注水后，核电厂反应堆压力容器（RPV）下封头熔池达到最大衰变热的时刻比无堆内注水的工况延迟了约8h，且衰变热也相应减小了近30%。有堆内注水的工况比无堆内注水的工况热流密度大幅减小，峰值减小了30%以上，大大增加了IVR安全裕量。计算结果表明在堆内注水结束后，即使堆芯熔化进程重启，堆内熔融物总量没有因为堆内注水而减小，但因为时间进程大大延缓，因此衰变热也大大降低，所以减小了熔池传热的热流密度，安全裕量从10%增加至50%左右。本实施例能够显著提高反应堆的安全性能。
本说明书中各个实施例采用递进的方式描述，每个实施例重点说明的都是与其他实施例的不同之处，各个实施例之间相同相似部分互相参见即可。对于实施例公开的装置而言，由于其与实施例公开的方法相对应，所以描述的比较简单，相关之处参见方法部分说明即可。
专业人员还可以进一步意识到，结合本文中所公开的实施例描述的各示例的单元及算法步骤，能够以电子硬件、计算机软件或者二者的结合来实现，为了清楚地说明硬件和软件的可互换性，在上述说明中已经按照功能一般性地描述了各示例的组成及步骤。这些功能究竟以硬件还是软件方式来执行，取决于技术方案的特定应用和设计约束条件。专业技术人员可以对每个特定的应用来使用不同方法来实现所描述的功能，但是这种实现不应认为超出本发明的范围。
结合本文中所公开的实施例描述的方法或算法的步骤可以直接用硬件、处理器执行的软件模块，或者二者的结合来实施。软件模块可以置于随机存储器（RAM）、内存、只读存储器（ROM）、电可编程ROM、电可擦除可编程ROM、寄存器、硬盘、可移动磁盘、CD-ROM、或技术领域内所公知的任意其它形式的存储介质中。
以上实施例只为说明本发明的技术构思及特点，其目的在于让熟悉此项技术的人士能够了解本发明的内容并据此实施，并不能限制本发明的保护范围。凡跟本发明权利要求范围所做的均等变化与修饰，均应属于本发明权利要求的涵盖范围。

Claims (16)

  1. 一种核电站熔融物堆内滞留系统，其特征在于，包括堆外注水系统和堆内注水系统，所述堆外注水系统用于在反应堆（10）严重事故阶段向反应堆（10）的堆外注水，所述堆内注水系统用于在反应堆（10）严重事故阶段向反应堆（10）的堆内注水，所述堆内注水系统使用的水为含硼水。
  2. 根据权利要求1所述的核电站熔融物堆内滞留系统，其特征在于，所述堆内注水系统包括用于存储惰性气体和含硼水的高压水箱（20），所述高压水箱（20）的出水口通过管道连接第一止回阀V9的第一端，所述第一止回阀V9的第二端通过管道连接至反应堆（10）的堆内；
    所述高压水箱（20）内惰性气体的压强大于标准大气压，当反应堆（10）的堆内压强小于所述第一止回阀V9的第一端的压强时，所述高压水箱（20）开始向反应堆（10）的堆内注入含硼水，且在含硼水注入完毕后注入所述高压水箱（20）内的惰性气体。
  3. 根据权利要求2所述的核电站熔融物堆内滞留系统，其特征在于，所述高压水箱（20）的出水口和所述第一止回阀V9之间管道上设置有电动阀V8；
    所述电动阀V8处于常开状态，所述电动阀V8在反应堆（10）维修时关闭。
  4. 根据权利要求2所述的核电站熔融物堆内滞留系统，其特征在于，所述高压水箱（20）内惰性气体的压强范围为0.1MPa至2MPa。
  5. 根据权利要求2所述的核电站熔融物堆内滞留系统，其特征在于，所述高压水箱（20）的出水口至所述第一止回阀V9的第一端之间管道和所述第一止回阀V9的第二端至反应堆（10）的堆内之间管道的管道内径范围为50mm至80mm。
  6. 根据权利要求2所述的核电站熔融物堆内滞留系统，其特征在于，所述高压水箱（20）的最低点高于反应堆（10）的最高点。
  7. 根据权利要求2所述的核电站熔融物堆内滞留系统，其特征在于，还包括与所述高压水箱（20）的入水口管道连接的换料水箱（30），所述换料水箱（30）内存储有含硼水；
    所述高压水箱（20）的入水口和所述换料水箱（30）之间的管道上设置有电动泵P1和电动阀V11，所述换料水箱（30）用于向所述高压水箱（20）内补充含硼水。
  8. 根据权利要求1至7任一项所述的核电站熔融物堆内滞留系统，其特征在于，所述堆外注水系统包括高位常压水箱（40）、电动阀V4和第二止回阀V6，所述高位常压水箱（40）的最低点高于反应堆（10）的最高点；
    所述高位常压水箱（40）的底部出水口通过管道连接所述电动阀V4的第一端，所述电动阀V4的第二端通过管道连接所述第二止回阀V6的第一端，所述第二止回阀V6的第二端通过管道连接至反应堆（10）的堆外；
    所述电动阀V4在反应堆（10）正常工作时处于关闭状态，在反应堆（10）堆芯出口温度超过650℃后开启，开启后所述高位常压水箱（40）内的水注入反应堆（10）的堆外。
  9. 根据权利要求8所述的核电站熔融物堆内滞留系统，其特征在于，还包括备用电动阀V5，所述备用电动阀V5的第一端通过管道连接所述电动阀V4的第一端，所述备用电动阀V5的第二端通过管道连接所述电动阀V4的第二端；
    所述备用电动阀V5在反应堆（10）正常工作时处于关闭状态，在反应堆（10）堆芯出口温度超过650℃后首先开启所述电动阀V4，若所述电动阀V4开启失败后开启所述备用电动阀V5，所述高位常压水箱（40）内的水注入反应堆（10）的堆外。
  10. 根据权利要求8所述的核电站熔融物堆内滞留系统，其特征在于，还包括设置在所述第二止回阀V6的第二端和反应堆（10）的堆外之间管道上的截止阀V7，所述截止阀V7用于在发生误注水时终止注水。
  11. 根据权利要求8所述的核电站熔融物堆内滞留系统，其特征在于，所述高位常压水箱（40）包括至少两个底部出水口，每个底部出水口通过管道连接至所述电动阀V4的第一端，每个底部出水口的管道上对应设置一个电动阀，所有所述电动阀在反应堆（10）正常工作时处于关闭状态，在反应堆（10）堆芯出口温度超过650℃后开启。
  12. 根据权利要求11所述的核电站熔融物堆内滞留系统，其特征在于，每个所述底部出水口对应管道的管口延伸至所述高位常压水箱（40）的内部，每个所述底部出水口对应管道的管口的延伸高度不同，且延伸高度最高的管道的管口之上的水量等于反应堆（10）的堆坑自由容积。
  13. 根据权利要求12所述的核电站熔融物堆内滞留系统，其特征在于，所述高位常压水箱（40）包括3个底部出水口，第一底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V1，第二底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V2，第三底部出水口至所述电动阀V4的第一端之间管道上设置有电动阀V3，所述电动阀V1、所述电动阀V2和所述电动阀V3在反应堆（10）正常工作时处于关闭状态，在反应堆（10）堆芯出口温度超过650℃后开启；
    所述第一底部出水口的对应管道在高位常压水箱（40）内无延伸，所述第三底部出水口的对应管道在高位常压水箱（40）内的延伸高度高于所述第二底部出水口的对应管道在所述高位常压水箱（40）内的延伸高度，且所述第三底部出水口对应管道的管口之上的水量等于反应堆（10）的堆坑自由容积。
  14. 根据权利要求8所述的核电站熔融物堆内滞留系统，其特征在于，还包括与所述高位常压水箱（40）的入水口管道连接的换料水箱（30），所述高位常压水箱（40）的入水口和所述换料水箱（30）之间的管道上设置有电动泵P1和电动阀V10，所述换料水箱（30）用于向所述高位常压水箱（40）内补充水。
  15. 根据权利要求14所述的核电站熔融物堆内滞留系统，其特征在于，所述高位常压水箱（40）内的水为含硼水；
    所述换料水箱（30）内的水为含硼水。
  16. 根据权利要求1所述的核电站熔融物堆内滞留系统，其特征在于，系统中所有电动阀采用不间断电源供电。
PCT/CN2020/112845 2020-09-01 2020-09-01 一种核电站熔融物堆内滞留系统 WO2022047622A1 (zh)

Priority Applications (2)

Application Number Priority Date Filing Date Title
PCT/CN2020/112845 WO2022047622A1 (zh) 2020-09-01 2020-09-01 一种核电站熔融物堆内滞留系统
CN202080106605.2A CN116391238A (zh) 2020-09-01 2020-09-01 一种核电站熔融物堆内滞留系统

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
PCT/CN2020/112845 WO2022047622A1 (zh) 2020-09-01 2020-09-01 一种核电站熔融物堆内滞留系统

Publications (1)

Publication Number Publication Date
WO2022047622A1 true WO2022047622A1 (zh) 2022-03-10

Family

ID=80492098

Family Applications (1)

Application Number Title Priority Date Filing Date
PCT/CN2020/112845 WO2022047622A1 (zh) 2020-09-01 2020-09-01 一种核电站熔融物堆内滞留系统

Country Status (2)

Country Link
CN (1) CN116391238A (zh)
WO (1) WO2022047622A1 (zh)

Citations (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH04109197A (ja) * 1989-09-19 1992-04-10 Mitsubishi Heavy Ind Ltd 加圧水型原子炉の炉心崩壊熱除去装置
CN102903404A (zh) * 2012-08-20 2013-01-30 中国核电工程有限公司 一种核电站能动与非能动结合的堆芯剩余热量排出系统
CN103632736A (zh) * 2012-08-20 2014-03-12 中国核动力研究设计院 一种核电站堆腔注水冷却系统
WO2015175878A1 (en) * 2014-05-15 2015-11-19 Holtec International An improved passively-cooled spent nuclear fuel pool system
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
CN205751540U (zh) * 2016-05-18 2016-11-30 中广核研究院有限公司 核电站严重事故缓解系统
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
CN111128414A (zh) * 2019-12-31 2020-05-08 中国核动力研究设计院 一种核电厂能动与非能动相结合的安全系统及其方法

Patent Citations (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH04109197A (ja) * 1989-09-19 1992-04-10 Mitsubishi Heavy Ind Ltd 加圧水型原子炉の炉心崩壊熱除去装置
CN102903404A (zh) * 2012-08-20 2013-01-30 中国核电工程有限公司 一种核电站能动与非能动结合的堆芯剩余热量排出系统
CN103632736A (zh) * 2012-08-20 2014-03-12 中国核动力研究设计院 一种核电站堆腔注水冷却系统
WO2015175878A1 (en) * 2014-05-15 2015-11-19 Holtec International An improved passively-cooled spent nuclear fuel pool system
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
CN205751540U (zh) * 2016-05-18 2016-11-30 中广核研究院有限公司 核电站严重事故缓解系统
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
CN111128414A (zh) * 2019-12-31 2020-05-08 中国核动力研究设计院 一种核电厂能动与非能动相结合的安全系统及其方法

Also Published As

Publication number Publication date
CN116391238A (zh) 2023-07-04

Similar Documents

Publication Publication Date Title
US7983376B2 (en) Boiling water nuclear reactor and emergency core cooling system of the same
US6795518B1 (en) Integral PWR with diverse emergency cooling and method of operating same
CN201689688U (zh) 堆芯冷却、堆腔充水及安全壳热量导出的系统
KR100300889B1 (ko) 가압수형원자로및증기발생기관의누출을완화시키는방법
KR101242746B1 (ko) 원자력 발전소의 격납건물 외부 통합피동안전계통 시스템
US20080219395A1 (en) Nuclear power plant using nanoparticles in emergency situations and related method
KR890001251B1 (ko) 가압 수로형 원자로용 급속냉각장치
CN111081399B (zh) 核电厂应急堆芯冷却系统
CN109903863B (zh) 一种安全注入系统及核电系统
KR100813939B1 (ko) 안전보호용기를 구비한 일체형원자로의 피동형비상노심냉각설비
CN201788707U (zh) 一种用于保证核电站安全的安全系统
WO2022047622A1 (zh) 一种核电站熔融物堆内滞留系统
KR20060020756A (ko) 다양한 비상냉각설비를 갖춘 일체형 가압 경수로 및 그운전방법
CN113661547B (zh) 核电厂严重事故的应对安全系统及其控制方法
CN103426484A (zh) 一种防止反应堆堆内熔融物熔损压力容器的方法以及用于实施这种方法的系统
CN113972016B (zh) 核电厂安全壳外失水事故应对方法、装置、设备及介质
CN212230087U (zh) 一种利用浮球阀控制的池式反应堆余热排出系统
JP2548838B2 (ja) 加圧水型原子炉の炉心崩壊熱除去装置
JPS6375691A (ja) 自然循環型原子炉
CN108597630B (zh) 一种核电厂全压非能动重力注入系统
CN111899902A (zh) 一种利用浮球阀控制的池式反应堆余热排出系统
CN220651664U (zh) 非能动安全系统及基于atf燃料的反应堆
CN216457919U (zh) 恒定液位组合氧化氮压缩机组分离器系统
CN103426485B (zh) 一种防止反应堆堆内熔融物熔损压力容器的方法以及用于实施这种方法的系统
CN210692106U (zh) 一种核电站卸压系统

Legal Events

Date Code Title Description
121 Ep: the epo has been informed by wipo that ep was designated in this application

Ref document number: 20951870

Country of ref document: EP

Kind code of ref document: A1

NENP Non-entry into the national phase

Ref country code: DE

32PN Ep: public notification in the ep bulletin as address of the adressee cannot be established

Free format text: NOTING OF LOSS OF RIGHTS PURSUANT TO RULE 112(1) EPC (EPO FORM 1205A DATED 03/07/2023)

122 Ep: pct application non-entry in european phase

Ref document number: 20951870

Country of ref document: EP

Kind code of ref document: A1