WO2022126445A1 - 预防堆芯熔融物熔穿rpv的安全系统及安全控制方法 - Google Patents

预防堆芯熔融物熔穿rpv的安全系统及安全控制方法 Download PDF

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Publication number
WO2022126445A1
WO2022126445A1 PCT/CN2020/136914 CN2020136914W WO2022126445A1 WO 2022126445 A1 WO2022126445 A1 WO 2022126445A1 CN 2020136914 W CN2020136914 W CN 2020136914W WO 2022126445 A1 WO2022126445 A1 WO 2022126445A1
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WIPO (PCT)
Prior art keywords
pit
cooling water
core melt
pressure vessel
rpv
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PCT/CN2020/136914
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English (en)
French (fr)
Inventor
展德奎
夏少雄
陈鹏
符卉
吴梓杰
贺东钰
刘春容
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 CN202080106617.5A priority Critical patent/CN116368580A/zh
Priority to PCT/CN2020/136914 priority patent/WO2022126445A1/zh
Priority to EP20965463.1A priority patent/EP4250315A4/en
Publication of WO2022126445A1 publication Critical patent/WO2022126445A1/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
    • G21C9/00—Emergency protection arrangements structurally associated with the reactor, e.g. safety valves provided with pressure equalisation devices
    • G21C9/016—Core catchers
    • 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 and a safety control method for preventing core melt from melting through an RPV.
  • the core undergoes a process of water loss, heating and melting, and the molten core will collapse and fall to the lower head of the reactor pressure vessel (RPV). If the high-temperature core melt cannot be cooled effectively in time, the core melt may melt through the RPV, resulting in a large amount of radioactive material leakage.
  • the molten material retention (IVR) strategy was adopted, and a water injection system was set up in the reactor pit.
  • the natural circulation cooling method is different for RPV cooling.
  • the low-power nuclear power plant adopts the pool boiling method to cool the RPV, which takes away the sensible heat and decay heat of the core melt, and finally cools the core melt and stays under the reactor pressure vessel. head, ensuring the integrity of the pressure vessel.
  • the technical problem to be solved by the present invention is to provide a simple and reliable safety system for preventing core melt from penetrating RPV and a safety control method for preventing core melt from penetrating RPV.
  • the technical scheme adopted by the present invention to solve the technical problem is to provide a safety system for preventing the core melt from melting through the RPV, including a reactor pit, a cooling water pool, a cooling channel and a water injection pipeline for accommodating the reactor pressure vessel;
  • the cooling water pool is arranged on the periphery of the stack pit, the cooling channel is arranged at the bottom of the stack pit and communicated with the cooling water pool; the water injection pipeline is connected between the cooling water pool and the stack pit, The cooling water in the cooling water pool is passively injected into the pile pit through the water injection pipeline under the action of gravity.
  • a shielding and sealing layer is provided at the pit opening of the stack pit to seal the pit opening and leave a steam passage; the steam passage is located between the shielding and sealing layer and the outer wall of the reactor pressure vessel.
  • a specular reflection layer is provided on the inner peripheral wall surface and the inner bottom surface of the pile pit.
  • the safety system further comprises at least one anti-convection plate arranged on the inner peripheral wall of the pile pit;
  • the first end of the anti-convection plate is connected to the specular reflection layer, and the opposite second end is inclined upward and close to the outer wall surface of the reactor pressure vessel.
  • the interval between the second end of the anti-convection plate and the outer wall of the reactor pressure vessel is 40mm-60mm.
  • the water injection pipeline is provided with a control valve for controlling the on-off of the water injection pipeline.
  • the pile pit is made of stainless steel plate, and the side wall of the pile pit forms the inner ring side wall of the cooling water pool.
  • the bottom plate of the pile pit comprises two stainless steel plates spaced up and down, and the interlayer between the two stainless steel plates forms the cooling channel.
  • the present invention also provides a safety control method for preventing core melt from penetrating RPV, using any of the above safety systems for preventing core melt from penetrating RPV, the safety control method includes the following steps: In the event of an accident, before the core melt collapses to the lower head of the reactor pressure vessel, the cooling water in the cooling pool is passively injected into the pit under the action of gravity through the water injection line.
  • the cooling water in the reactor pit boils and evaporates after exchanging heat with the reactor pressure vessel, taking away the heat on the outer surface of the reactor pressure vessel, and at the same time transferring the heat to the cooling water pool through thermal convection and heat conduction.
  • the safety control method further includes the following steps: in an extreme working condition in which the reactor pressure vessel is melted through by the core melt, the cooling water in the cooling channel at the bottom of the pile pit is different from the cooling water entering the pile pit and located in the pile pit.
  • the cooling water below the reactor pressure vessel conducts heat exchange, cools the bottom of the reactor pressure vessel, takes away the heat of the core melt, and prevents the core melt from burning through the bottom plate of the reactor pit.
  • part of the water vapor in the stack pit is discharged to the outside through a steam channel at the pit opening; the discharged water vapor is condensed into water and then returned to the cooling water pool.
  • the safety control method further includes the following steps: under normal operating conditions, the cooling space between the reactor pit and the reactor pressure vessel forms a heat preservation space formed by relatively still air, and the heat of the reactor pressure vessel is conducted to the reactor in the pit. The air is then conducted to the cooling pool through the side wall of the pile pit.
  • the beneficial effects of the invention are as follows: in the case of serious accidents and other working conditions, without relying on an external AC power supply, the passive water injection is carried out using the gravitational potential energy into the pile pit, the system is simple and the reliability is high, and the low-power nuclear power constant core fusion reactor can be realized. Internal cooling and retention, improve the safety of nuclear reactors, and reduce the risk of large-scale release of radioactive materials.
  • FIG. 1 is a schematic top-view structural diagram of a safety system for preventing core melt from melting through an RPV according to an embodiment of the present invention
  • FIG. 2 is a schematic cross-sectional structural diagram of a safety system for preventing core melt from melting through an RPV according to an embodiment of the present invention.
  • a safety system for preventing core melt from melting through an RPV includes a reactor pit 10 for accommodating a reactor pressure vessel 100 (RPV), a cooling water pool 20 , and a cooling channel 30 and water injection line 40 .
  • the reactor pressure vessel 100 is suspended in the reactor pit 10, and a space is left between the outer surface (including the outer wall and the bottom surface) of the reactor pressure vessel 100 and the reactor pit 10 to form a cooling space for injecting cooling water or circulating air therein.
  • the cooling water pool 20 is arranged on the periphery of the reactor pit 10 , and can provide cooling water for the cooling space, and can also perform heat exchange with the cooling water in the reactor pit 10 to realize the cooling of the reactor pressure vessel 100 .
  • the cooling channel 30 is arranged at the bottom of the stack pit 10 and communicated with the cooling water pool 20, so that the cooling channel 30 is also filled with cooling water, so that heat exchange can be performed with the cooling water at the bottom of the stack pit 10, and the bottom of the reactor pressure vessel 100 can be cooled. Cooling to realize the double-layer core melt retention (IVR) concept.
  • IVR double-layer core melt retention
  • the water injection line 40 is connected between the cooling water pool 20 and the heap pit 10.
  • the cooling water in the cooling water pool 20 is passively injected into the heap pit 10 through the water injection line 40 under the action of gravity, without relying on an external AC power supply, and has high reliability.
  • the water injection line 40 is preferably connected between the upper end of the cooling water pool 20 and the upper end of the stack pit 10, so that the cooling water can be passively injected into the stack pit 10 and the submerged water level is above the active section of the core.
  • the water injection pipeline 40 is provided with a control valve 41 for controlling the on-off of the water injection pipeline 40 .
  • the control valve may be an electric valve.
  • the pile pit 10 is made of stainless steel plates, so that the side walls and the bottom plate of the pile pit 10 are both formed of stainless steel plates. Dimensions such as the inner diameter and depth of the reactor pit 10 are set according to the requirements of the reactor pressure vessel 100 .
  • the side walls of the pile pit 10 form the inner ring side walls of the cooling water pool 20 .
  • the outer ring side wall of the cooling water pool 20 is also made of stainless steel.
  • the bottom plate of the stack pit 10 includes two stainless steel plates spaced up and down, and the interlayer between the two stainless steel plates forms a cooling channel 30; the two ends of the cooling channel 30 can be directly opened to communicate with the cooling water pool 20, or a vertical plate can be provided for support And through holes are opened on the vertical plate to realize the circulation of cooling water.
  • the reactor pressure vessel 100 is arranged in the reactor pit 10 and occupies most of the space in the reactor pit 10.
  • the volume of the cooling space between the reactor pressure vessel 100 and the inner wall of the reactor pit 10 is small, so that the cooling space can be replaced in a short time. Fill up with cooling water and flood the pit 10. After the cooling water enters the reactor pit 10 , the reactor pressure vessel 100 is directly cooled by the pool boiling method, and the cooling water of the core melt in the lower head of the reactor pressure vessel 100 is taken away to maintain the integrity of the reactor pressure vessel 100 .
  • a shielding and sealing layer 50 is provided at the pit opening of the heap pit 10 to seal the pit opening and leave a steam channel 200; the steam channel 200 is located between the shielding and sealing layer 50 and the reactor pressure. between the outer walls of the container 100 .
  • the steam passage 200 communicates the cooling space in the heap pit 10 with the large space outside the heap pit 10 (containment large space), and the steam used in the heap pit 10 is discharged from the heap pit 10 through.
  • a specular reflection layer 60 is provided on the inner peripheral wall surface (ie, the inner wall surface of the side wall) and the inner bottom surface (ie, the upper surface of the bottom plate) of the pile pit 10 to reduce heat dissipation caused by thermal radiation.
  • the safety system of the present invention further includes at least one anti-convection plate 70 disposed on the inner peripheral wall of the pile pit 10 .
  • the anti-convection plate 70 has opposite first and second ends, the first end of the anti-convection plate 70 is connected to the mirror reflection layer 60, and the second end is inclined upward and close to the outer wall surface of the reactor pressure vessel 100, so as to resist convection
  • the plate 70 is arranged in an upwardly inclined manner within the stack pit 10 .
  • the arrangement of the anti-convection plate 70 can further reduce the air convection on the outer wall of the reactor pressure vessel 100, and at the same time, it can also ensure the overflow of steam when the outer wall of the RPV undergoes pool boiling after the reactor pit 10 is flooded with water under severe accident conditions.
  • anti-convection plates 70 When a plurality of anti-convection plates 70 are provided, they are distributed in the stack pit 10 at intervals along the height direction of the stack pit 10 .
  • the interval between the second end of the anti-convection plate 70 and the outer wall surface of the reactor pressure vessel 100 is 40mm-60mm.
  • the cooling water pool 20 is a large cooling water pool, surrounding the periphery of the heap pit 10, the volume of the cooling water pool 20 is greater than 500 m 3 , and the wall thickness of the cooling water pool 20 is 20-40 mm; The free volume (ie cooling space volume) is 7m 3 .
  • the heat can be directly transferred to the cooling water pool 20 .
  • a small amount of steam may be discharged into the containment through the steam channel 200 at the top of the pit 10 .
  • the distance between the reactor pressure vessel 100 and the bottom of the reactor pit 10 is relatively small, and the distance is about 200 mm. Under extreme conditions, assuming that the core melt melts through the RPV, the melt will reset to the bottom of the RPV. At this time, due to the presence of the cooling water pool 20 and the cold water channel 30 at the bottom of the stack pit 10, secondary core melt will be formed. The material cooling and retaining device is used to finally cool the core melt and retain it inside the reactor pit 10 .
  • the safety control method for preventing the core melt from penetrating the RPV of the present invention is realized by the above-mentioned safety system, and the safety control method may include the following steps: when a serious accident occurs, when the core melt collapses to the bottom of the reactor pressure vessel 100 Before the head, the cooling water in the cooling water pool 20 is passively injected into the heap pit 10 under the action of gravity through the water injection line 40 .
  • the initial height of the water surface of the cooling water pool 20 is higher than the upper surface of the active section of the core, and the volume of the higher part can satisfy that the submerged water level after water injection is still above the active section of the core.
  • the cooling water in the reactor pit 10 exchanges heat with the reactor pressure vessel 100 and then boils and evaporates, taking away the heat from the outer surface of the reactor pressure vessel 100 , and also conducts the heat to the cooling pool 20 through thermal convection and heat conduction.
  • Part of the water vapor (a small amount) in the stack pit 10 is discharged to the outside through the steam channel 200 at the pit opening; the discharged water vapor is condensed into water and then returned to the cooling water pool 20, so as to implement the melt pressure vessel retention (IVR) Provides long-term stable heat sinks.
  • IVR melt pressure vessel retention
  • the cooling water in the cooling channel 30 at the bottom of the reactor pit 10 exchanges heat with the cooling water that enters the reactor pit 10 and is located below the reactor pressure vessel 100 .
  • the bottom of the reactor pressure vessel 100 is cooled to take away the heat of the core melt and prevent the core melt from burning through the bottom plate of the reactor pit 10 .
  • the cooling space between the reactor pit 10 and the reactor pressure vessel 100 forms a heat preservation space formed by relatively still air, and the heat of the reactor pressure vessel 100
  • the air conducted to the stack pit 10 is then conducted to the cooling water pool 20 through the side wall of the stack pit 10 .

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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

本发明公开了一种预防堆芯熔融物熔穿RPV的安全系统及安全控制方法，安全系统包括用于容置反应堆压力容器的堆坑、冷却水池、冷却通道以及注水管线；所述冷却水池设置在所述堆坑的外围，所述冷却通道设置在所述堆坑的底部并与所述冷却水池相连通；所述注水管线连接在所述冷却水池和堆坑之间，所述冷却水池内的冷却水在重力作用下通过所述注水管线以非能动方式注入所述堆坑。本发明的预防堆芯熔融物熔穿RPV的安全系统，满足在严重事故等工况下，在不依靠外部交流电源情况，利用重力势能往堆坑进行非能动注水，系统简单、可靠性高。

Description

预防堆芯熔融物熔穿RPV的安全系统及安全控制方法 技术领域
本发明涉及核电技术领域，尤其涉及一种预防堆芯熔融物熔穿RPV的安全系统及安全控制方法。
 背景技术
对于低功率核电厂（200-300MWe）来说，在发生严重事故后，堆芯经历失水、升温和熔化过程，熔化的堆芯会坍塌掉落至反应堆压力容器（RPV）下封头上。若高温的堆芯熔融物无法及时有效的被冷却，堆芯熔融物则可能熔穿RPV，最终导致放射性物质大量泄漏。为了避免RPV在严重事故条件下被熔穿，针对低功率核电厂的热功率较低的特点，采用了熔融物堆内滞留（IVR）策略，设置了堆坑注水系统，与高功率核电厂采用自然循环冷却方式进行RPV冷却的方式不同，低功率核电厂采用池式沸腾的方式进行冷却RPV，带走堆芯熔融物显热和衰变热，最终将堆芯熔融物冷却滞留在反应堆压力容器下封头中，确保压力容器的完整性。
现有大型压水堆核电厂（热功率600MWeu以上）大多采用堆芯熔融物冷却滞留策略实现严重事故下堆芯熔融物冷却堆内滞留，从而维持RPV在严重事故下的完整性。一般来说，大型核电厂在堆坑内部设计保温层流道，在严重事故事故后，通过非能动注水或者能动注水的方式向堆坑内注水；在淹没堆坑后，在堆坑内通过自然循环或者强迫循环方式驱动RPV外壁面冷却水进行对流换热。这种设计需要涉及金属保温层、保温层入水口、排气孔、循环流道等设计，但这种结构受保温层结构完整性影响较大，同时保温层入水口、排汽孔盖板的可靠性也对循环流动产生影响。此外，以上设备安装、造价、检修成本相对较高。
 技术问题
本发明要解决的技术问题在于，提供一种简单可靠的预防堆芯熔融物熔穿RPV的安全系统及预防堆芯熔融物熔穿RPV的安全控制方法。
 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：提供一种预防堆芯熔融物熔穿RPV的安全系统，包括用于容置反应堆压力容器的堆坑、冷却水池、冷却通道以及注水管线；
所述冷却水池设置在所述堆坑的外围，所述冷却通道设置在所述堆坑的底部并与所述冷却水池相连通；所述注水管线连接在所述冷却水池和堆坑之间，所述冷却水池内的冷却水在重力作用下通过所述注水管线以非能动方式注入所述堆坑。
优选地，所述堆坑的坑口处设有屏蔽封堵层，将所述坑口封堵并留有蒸汽通道；所述蒸汽通道位于所述屏蔽封堵层和反应堆压力容器的外壁之间。
优选地，所述堆坑的内周壁面及内底面上设有镜面反射层。
优选地，所述安全系统还包括设置在所述堆坑的内周壁面上的至少一个抗对流板；
所述抗对流板的第一端连接在所述镜面反射层上，相对的第二端倾斜朝上并靠近反应堆压力容器的外壁面。
优选地，所述抗对流板的第二端与所述反应堆压力容器的外壁面的间隔为40mm-60mm。
优选地，所述注水管线设有控制注水管线通断的控制阀门。
优选地，所述堆坑由不锈钢板制成，所述堆坑的侧壁形成所述冷却水池的内圈侧壁。
优选地，所述堆坑的底板包括上下间隔的两个不锈钢板，该两个不锈钢板之间的夹层形成所述冷却通道。
本发明还提供一种预防堆芯熔融物熔穿RPV的安全控制方法，采用以上任一项所述的预防堆芯熔融物熔穿RPV的安全系统，所述安全控制方法包括以下步骤：发生严重事故时，在堆芯熔融物坍塌至反应堆压力容器的下封头前，通过注水管线使冷却水池中的冷却水在重力作用下以非能动方式注入堆坑。
优选地，所述堆坑内的冷却水与反应堆压力容器换热后沸腾蒸发，带走反应堆压力容器外表面的热量，同时还通过热对流及热传导方式将热量传导至冷却水池。
优选地，所述安全控制方法还包括以下步骤：在反应堆压力容器被堆芯熔融物熔穿的极端工况下，所述堆坑底部的冷却通道内的冷却水与进入所述堆坑并位于反应堆压力容器下方的冷却水进行热交换，对反应堆压力容器底部进行冷却，带走堆芯熔融物的热量，防止堆芯熔融物烧穿堆坑的底板。
优选地，所述堆坑内部分水蒸汽通过其坑口处的蒸汽通道向外排出；排出的水蒸汽冷凝成水后回流至所述冷却水池内。
优选地，所述安全控制方法还包括以下步骤：在正常运行条件下，堆坑与反应堆压力容器之间的冷却空间形成由相对静止空气形成的保温空间，反应堆压力容器的热量传导给堆坑内的空气，再通过堆坑的侧壁传导给冷却水池。
 有益效果
本发明的有益效果：满足在严重事故等工况下，在不依靠外部交流电源情况，利用重力势能往堆坑进行非能动注水，系统简单、可靠性高，实现低功率核电常堆芯熔融堆内冷却滞留，提高核反应堆安全性，降低放射性物质大规模释放的风险。
 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预防堆芯熔融物熔穿RPV的安全系统的俯视结构示意图；
图2是本发明一实施例的预防堆芯熔融物熔穿RPV的安全系统的剖面结构示意图。
 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、2所示，本发明一实施例的预防堆芯熔融物熔穿RPV的安全系统，包括用于容置反应堆压力容器100（RPV）的堆坑10、冷却水池20、冷却通道30以及注水管线40。
反应堆压力容器100悬空在堆坑10内，其外表面（包括外壁面和底面）与堆坑10之间留有间隔，形成用于注入冷却水或空气在其中流通的冷却空间。冷却水池20设置在堆坑10的外围，可以为冷却空间提供冷却水，同时也可以与堆坑10内的冷却水进行热交换，实现反应堆压力容器100的冷却。冷却通道30设置在堆坑10的底部并与冷却水池20相连通，使得冷却通道30内也充满冷却水，从而可与位于堆坑10底部的冷却水进行热交换，实现反应堆压力容器100底部的冷却，实现双层堆芯熔融物滞留（IVR）理念。
注水管线40连接在冷却水池20和堆坑10之间，冷却水池20内的冷却水在重力作用下通过注水管线40以非能动方式注入堆坑10，不依靠外部交流电源，可靠性高。注水管线40优选连接在冷却水池20的上端和堆坑10的上端之间，以使冷却水可以非能动注入堆坑10并使淹没水位在堆芯活性段以上。
注水管线40上设有控制阀门41，用于控制该注水管线40的通断。控制阀门可以是电动阀。
如图2所示，本实施例中，堆坑10由不锈钢板制成，使得堆坑10的侧壁及底板均是不锈钢板构成。堆坑10的内径、深度等尺寸均根据反应堆压力容器100需求进行设置。堆坑10的侧壁形成冷却水池20的内圈侧壁。同样地，冷却水池20的外圈侧壁也有不锈钢板制成。堆坑10的底板包括上下间隔的两个不锈钢板，该两个不锈钢板之间的夹层形成冷却通道30；冷却通道30的两端可直接开放而连通冷却水池20，或者设有竖板实现支撑并在竖板上开设通孔实现冷却水的流通。
反应堆压力容器100设置在堆坑10内，占据了堆坑10内的大部分空间，反应堆压力容器100与堆坑10内壁之间的冷却空间体积较小，这样使得可以在短时间内将冷却空间注满冷却水，淹没堆坑10。冷却水进入堆坑10后直接通过池式沸腾方式冷却反应堆压力容器100，将反应堆压力容器100下封头内堆芯熔融物冷却水带走，维持反应堆压力容器100完整性。
为了减少堆坑10内部空气对流导致热量耗散，堆坑10的坑口处设有屏蔽封堵层50，将坑口封堵并留有蒸汽通道200；蒸汽通道200位于屏蔽封堵层50和反应堆压力容器100的外壁之间。蒸汽通道200将堆坑10内的冷却空间与堆坑10外部的大空间（安全壳大空间）相连通，且用于堆坑10内的蒸汽通过排出堆坑10。
堆坑10的内周壁面（即侧壁的内壁面）及内底面（即底板的上表面）上设有镜面反射层60，用于减少热辐射导致热量耗散。
进一步地，本发明的安全系统还包括设置在堆坑10的内周壁面上的至少一个抗对流板70。抗对流板70具有相对的第一端和第二端，抗对流板70的第一端连接在镜面反射层60上，第二端倾斜朝上并靠近反应堆压力容器100的外壁面，使得抗对流板70在堆坑10内呈向上倾斜设置。
抗对流板70的设置可以进一步减少反应堆压力容器100外壁面空气对流，同时还可以保证在严重事故条件下实施堆坑10注水淹没后，RPV外壁面发生池式沸腾时蒸汽的溢出。
抗对流板70设有多个时，在堆坑10内沿堆坑10的高度方向间隔分布。
作为选择，抗对流板70的第二端与反应堆压力容器100的外壁面的间隔为40mm-60mm。
在一种选择性实施方式中，冷却水池20为大型冷却水池，围绕在堆坑10的外围，冷却水池20的容积大于500m 3，冷却水池20的壁厚为20-40mm；堆坑10内的自由体积（即冷却空间体积）为7m 3。堆坑10内冷却水在被RPV加热后，可以直接将热量传到给冷却水池20。少量的蒸汽可以通过堆坑10顶部的蒸汽通道200排入到安全壳中。
反应堆压力容器100与堆坑10底部间距较小，间距约200mm。在极端工况下，假设堆芯熔融物熔穿RPV，熔融物会重置到RPV的底部，此时由于冷却水池20以及堆坑10底部的冷水通道30的存在，会形成二次堆芯熔融物冷却滞留装置，将堆芯熔融物最终冷却滞留在堆坑10内部。
本发明的预防堆芯熔融物熔穿RPV的安全控制方法，通过上述的安全系统实现，该安全控制方法可包括以下步骤：发生严重事故时，在堆芯熔融物坍塌至反应堆压力容器100的下封头前，通过注水管线40使冷却水池20中的冷却水在重力作用下以非能动方式注入堆坑10。
冷却水池20水面初始高度高于堆芯活性段上表面，且高出部分的体积可以满足注水后的淹没水位仍在堆芯活性段以上。
堆坑10内的冷却水与反应堆压力容器100换热后沸腾蒸发，带走反应堆压力容器100外表面的热量，同时还通过热对流及热传导方式将热量传导至冷却水池20。堆坑10内部分水蒸汽（少量）通过其坑口处的蒸汽通道200向外排出；排出的水蒸汽冷凝成水后回流至冷却水池20内，从而为熔融物压力容器内滞留（IVR）的实施提供长期稳定热阱。
在反应堆压力容器100被堆芯熔融物熔穿的极端工况下，堆坑10底部的冷却通道30内的冷却水与进入堆坑10并位于反应堆压力容器100下方的冷却水进行热交换，对反应堆压力容器100底部进行冷却，带走堆芯熔融物的热量，防止堆芯熔融物烧穿堆坑10的底板。
此外，结合本发明的安全系统，在正常运行条件下（不发生严重事故），堆坑10与反应堆压力容器100之间的冷却空间形成由相对静止空气形成的保温空间，反应堆压力容器100的热量传导给堆坑10的空气，再通过堆坑10的侧壁传导给冷却水池20。
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 (13)

  1. 一种预防堆芯熔融物熔穿RPV的安全系统，其特征在于，包括用于容置反应堆压力容器的堆坑、冷却水池、冷却通道以及注水管线；
    所述冷却水池设置在所述堆坑的外围，所述冷却通道设置在所述堆坑的底部并与所述冷却水池相连通；所述注水管线连接在所述冷却水池和堆坑之间，所述冷却水池内的冷却水在重力作用下通过所述注水管线以非能动方式注入所述堆坑。
  2. 根据权利要求1所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述堆坑的坑口处设有屏蔽封堵层，将所述坑口封堵并留有蒸汽通道；所述蒸汽通道位于所述屏蔽封堵层和反应堆压力容器的外壁之间。
  3. 根据权利要求1所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述堆坑的内周壁面及内底面上设有镜面反射层。
  4. 根据权利要求3所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述安全系统还包括设置在所述堆坑的内周壁面上的至少一个抗对流板；
    所述抗对流板的第一端连接在所述镜面反射层上，相对的第二端倾斜朝上并靠近反应堆压力容器的外壁面。
  5. 根据权利要求4所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述抗对流板的第二端与所述反应堆压力容器的外壁面的间隔为40mm-60mm。
  6. 根据权利要求1所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述注水管线设有控制注水管线通断的控制阀门。
  7. 根据权利要求1-6任一项所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述堆坑由不锈钢板制成；所述堆坑的侧壁形成所述冷却水池的内圈侧壁。
  8. 根据权利要求7所述的预防堆芯熔融物熔穿RPV的安全系统，其特征在于，所述堆坑的底板包括上下间隔的两个不锈钢板，该两个不锈钢板之间的夹层形成所述冷却通道。
  9. 一种预防堆芯熔融物熔穿RPV的安全控制方法，其特征在于，采用权利要求1-8任一项所述的预防堆芯熔融物熔穿RPV的安全系统，所述安全控制方法包括以下步骤：发生严重事故时，在堆芯熔融物坍塌至反应堆压力容器的下封头前，通过注水管线使冷却水池中的冷却水在重力作用下以非能动方式注入堆坑。
  10. 根据权利要求9所述的预防堆芯熔融物熔穿RPV的安全控制方法，其特征在于，所述堆坑内的冷却水与反应堆压力容器换热后沸腾蒸发，带走反应堆压力容器外表面的热量，同时还通过热对流及热传导方式将热量传导至冷却水池。
  11. 根据权利要求10所述的预防堆芯熔融物熔穿RPV的安全控制方法，其特征在于，还包括以下步骤：在反应堆压力容器被堆芯熔融物熔穿的极端工况下，所述堆坑底部的冷却通道内的冷却水与进入所述堆坑并位于反应堆压力容器下方的冷却水进行热交换，对反应堆压力容器底部进行冷却，带走堆芯熔融物的热量，防止堆芯熔融物烧穿堆坑的底板。
  12. 根据权利要求10所述的预防堆芯熔融物熔穿RPV的安全控制方法，其特征在于，所述堆坑内部分水蒸汽通过其坑口处的蒸汽通道向外排出；排出的水蒸汽冷凝成水后回流至所述冷却水池内。
  13. 根据权利要求9-12任一项所述的预防堆芯熔融物熔穿RPV的安全控制方法，其特征在于，还包括以下步骤：在正常运行条件下，堆坑与反应堆压力容器之间的冷却空间形成由相对静止空气形成的保温空间，反应堆压力容器的热量传导给堆坑内的空气，再通过堆坑的侧壁传导给冷却水池。
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