WO2017045159A1 - 核电站反应堆压力容器与屏蔽墙的组合结构 - Google Patents

核电站反应堆压力容器与屏蔽墙的组合结构 Download PDF

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WO2017045159A1
WO2017045159A1 PCT/CN2015/089758 CN2015089758W WO2017045159A1 WO 2017045159 A1 WO2017045159 A1 WO 2017045159A1 CN 2015089758 W CN2015089758 W CN 2015089758W WO 2017045159 A1 WO2017045159 A1 WO 2017045159A1
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pressure vessel
reactor pressure
pit
nuclear power
power plant
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PCT/CN2015/089758
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English (en)
French (fr)
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周媛霞
王庆礼
程浩
彭国胜
刘永
曹涛
陈兴
许晨德
黄威
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中广核工程有限公司
中国广核集团有限公司
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Priority to PCT/CN2015/089758 priority Critical patent/WO2017045159A1/zh
Publication of WO2017045159A1 publication Critical patent/WO2017045159A1/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
    • 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

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  • the invention belongs to the field of severe accident response of nuclear power plants, and more particularly, the invention relates to a combined structure of a nuclear power plant reactor pressure vessel and a shielding wall capable of coping with severe accident conditions.
  • the reactor core After a serious accident at the pressurized water reactor nuclear power plant, if the internal cooling means fails, the reactor core will gradually melt and collapse and fall into the bottom of the reactor pressure vessel. At this time, if effective external cooling is not obtained, the reactor pressure vessel will be melted through, causing the melt retention in the reactor to fail, the core melt entering the reactor pit, and then MCCI reaction with the concrete floor (melt and Concrete reacts) and releases hazardous combustible gases, causing a significant deterioration in severe accident conditions.
  • some of the disclosed pressurized water reactor nuclear power plants use the external cooling method to cool the reactor pressure vessel, that is, to inject water into the pit under severe accident conditions, so that the reactor pressure vessel is immersed in the pit water body and passed
  • the outer wall cooling mode ensures that the reactor pressure vessel is not melted through, so that the melt in the stack can be stably retained in the reactor pressure vessel.
  • the cooling medium In order to prevent the reactor pressure vessel from penetrating due to insufficient critical heat flux density, the cooling medium must flow through the outer wall of the reactor pressure vessel at a large flow rate. Therefore, it is necessary to provide a suitable flow passage for the external cooling medium of the reactor pressure vessel. In order to make the external cooling medium produce a strong natural circulation effect.
  • a suitable gap is set between the outer wall of the reactor pressure vessel and its thermal insulation layer, and a steam outlet and a water inlet are respectively arranged at the upper and lower portions of the thermal insulation layer, and the upper and lower water inlets are passively operated under severe accident conditions.
  • the principle is automatically turned on to provide an upflow path for off-cooling. Due to the small amount of steam in the steam-water mixture discharged from the upper steam outlet (about 1%), the main component is saturated water, and only try to make this part of the saturated water return to the bottom of the pile with a small pressure drop. In order to form a strong and stable natural circulation effect of extra-cooling. Experiments and analysis prove that increasing the natural circulation flow of the external cooling is beneficial to increase the critical heat flux density, so that the cooling capacity outside the reactor can be better ensured.
  • the special structural design of the reactor pressure vessel support ring provides a natural circulation return flow path for the external cooling of the reactor, but some of the nuclear power plant reactor support rings are higher than the heap overflow standard due to the seismic design requirements. High, it is not possible to provide a return flow path for extra-cooling. To this end, it is necessary to adjust the design of the pit structure to solve the problem of the return flow path of the external cooling. In addition, the pile pit structure is compact, and its functional requirements are many. It is necessary to reasonably integrate the functional design under normal operation and serious accidents, and solve the problem of the return flow path of the external cooling in this framework.
  • the object of the present invention is to provide a combined structure of a nuclear power plant reactor pressure vessel and a shield wall to ensure that the cooling medium can form a stable natural circulation under severe accident conditions, and more effectively cool the reactor pressure vessel to ensure the reactor pressure vessel. Integrity.
  • the present invention provides a combined structure of a nuclear power plant reactor pressure vessel and a shield wall, which comprises a pit shield wall, a pit surrounded by a pit shield wall, a reactor pressure vessel and a set in the pit.
  • a nuclear power plant reactor pressure vessel which comprises a pit shield wall, a pit surrounded by a pit shield wall, a reactor pressure vessel and a set in the pit.
  • the insulation layer outside the reactor pressure vessel between the insulation layer and the reactor pressure vessel, between the insulation layer and the shield wall of the pit, there are flow passages running up and down;
  • the lower part of the shield wall of the pit is provided with a pit water injection pipeline;
  • the upper inner wall of the trench shielding wall is provided with a reserved space, and the backflow shielding wall is provided with a return flow path connecting the receiving space with the bottom space of the pit.
  • the return flow path includes a bottom outlet for connecting the bottom space of the pit, and the bottom outlet is provided with an openable barrier; under normal working conditions, the blocking member closes the bottom outlet of the return flow path to prevent the normal air supply from the return flow.
  • Flow passage bypass under severe accident conditions, the blocking member automatically opens according to the passive principle, so that the return flow passage is unblocked, and the external cooling medium in the reserved space is smoothly returned to the bottom of the pile pit.
  • the flow path between the thermal insulation layer and the reactor pressure vessel is an ascending flow passage of the external cooling medium under severe accident; under severe accident conditions
  • the ascending flow path and the return flow path form a natural circulation loop under the influence of the difference in density of the medium, and the external pressure cooling of the reactor pressure vessel is continued.
  • the space for the shielding wall of the pit is connected to the external space of the pit through the inlet and outlet of the main pipe; in severe accident conditions, the rising channel
  • the vapor-liquid mixed medium is separated into steam and saturated water in the space, wherein the steam exits the pit through the inlet and outlet of the main pipe on the wall of the pit and takes away the heat of the pile, and the separated saturated water flows back through the return flow path. Go to the bottom of the pit.
  • the bottom of the thermal insulation layer is provided with a water injection hole, and a steam outlet is left between the top and the reactor pressure vessel; the upper steam outlet of the heat insulation layer and The lower water injection hole is normally closed, and is automatically opened by the passive principle under severe accident conditions, thereby forming an ascending flow path of the external cooling medium between the reactor pressure vessel and the insulation layer.
  • the water injection holes are all located below the lower head of the reactor pressure vessel.
  • the top inlet of the return flow channel is located at the lowest point of the space occupied by the pit shield wall.
  • the space of the trench shielding wall is located at the elevation of the nozzle section of the main pipeline of the reactor, and is annular.
  • the upper part of the reactor pressure vessel is connected with a plurality of main pipelines, and the main pipelines are correspondingly connected to the independent pipelines.
  • the tunnel is shielded from the pit; the inlet and outlet of the main pipe are larger than the cross-sectional area of the main pipe, thereby connecting the space to the external space of the pit.
  • the bottom wall of the annular accommodation space is provided with a support ring, and the support ring forms a support for the roots of all the main pipes, so that the reactor pressure vessel is suspended in the space.
  • the top inlet of the return flow channel is located on the bottom wall of the containment space outside the support ring.
  • the bottom wall of the main pipe inlet and outlet is higher than the bottom wall of the annular receiving space, and the top inlet of the return flow channel is located at the bottom of the annular receiving space. point.
  • the flow path between the thermal insulation layer and the shielding wall of the pit is a normal ventilation flow path, and is used as a flow of the ventilation of the pile under normal working conditions. aisle.
  • the combined structure of the nuclear power plant reactor pressure vessel and the shield wall of the present invention ensures that the saturated water entrained by the steam can smoothly flow back to the pile pit by setting a return flow passage in the shield wall of the pit, so that it can be in the pit A stable natural circulation is formed between the ascending flow path and the return flow path to ensure rapid and continuous cooling of the reactor pressure vessel under severe accident conditions, thereby effectively retaining the melt in the reactor pressure vessel.
  • FIG. 1 is a schematic view showing the combined structure of a nuclear power plant reactor pressure vessel and a shield wall according to the present invention.
  • the combined structure of the nuclear power plant reactor pressure vessel and the shield wall of the present invention includes a pit shield wall 10, a pit 20 surrounded by the pit shield wall 10, and a reactor pressure vessel 30 disposed in the pit 20.
  • the reactor pressure vessel 30 is connected at the upper portion with a plurality of main pipes 32 penetrating from the pile shield wall 10.
  • the pit shield wall 10 includes a side wall 12 and a bottom plate 14.
  • the side wall 12 has an annular receiving space 120 and a plurality of main pipe inlets and outlets 122 for communicating the receiving space 120 with the outer space of the pile pit on the inner wall of the upper portion; the receiving space 120 is located at the elevation of the main pipe connecting pipe section of the reactor.
  • the number and location of the main conduit inlets and outlets 122 correspond to the number and location of the main conduits 32.
  • a support ring 16 is mounted on the bottom wall of the accommodating space 120, and the support ring 16 supports the roots of all the main pipes 32 such that the reactor pressure vessel 30 is suspended and housed in the sump 20.
  • the main pipe inlet and outlet 122 Since the size of the main pipe inlet and outlet 122 is larger than the cross-sectional area of the main pipe 32, and the seal arrangement is not used therebetween, the main pipe inlet and outlet 122 becomes a passage for the pile 20 and the accommodation space 120 to communicate with the external space.
  • the bottom wall of the main pipe inlet and outlet 122 is higher than the bottom wall of the accommodating space 120, and the main pipe inlet and outlet 122 and the accommodating space 120 together form the steam outlet of the sump shielding wall 10.
  • the lower portion of the side wall 12 is sealed with a pit water injection line 18, and the outlet of the heap water injection line 18 is located near the bottom of the pile 20 for injecting water into the pile 20 under severe accident conditions.
  • the pit filling water pipeline 18 is connected with a water source such as a pit filling water tank and a refueling water tank in the safety shell, and water is injected into the pile pit 20 by means of passive, active or a combination of the two.
  • an insulating layer 34 is also provided outside thereof.
  • the heat insulating layer 34 is connected to the inner wall of the pit shield wall 10 by external support (not shown), and the inner wall thereof is not connected to the reactor pressure vessel 30. Therefore, there is a vertical connection between the inner wall of the heat insulating layer 34 and the outer wall of the reactor pressure vessel 30. Clearance.
  • the bottom of the heat insulating layer 34 is provided with a water injection hole 340 which is arranged laterally below the lower head of the reactor pressure vessel 30, and a steam outlet is left between the top of the heat insulating layer 34 and the reactor pressure vessel 30.
  • the upper steam outlet and the lower water injection hole 340 of the heat insulating layer 34 are normally closed, and are automatically opened by the passive principle under severe accident conditions, thereby forming an ascending flow passage A between the reactor pressure vessel 30 and the heat insulating layer 34 for The rise of the external cooling medium under severe accident conditions.
  • the outer wall of the heat insulating layer 34 is also not adhered to the inner wall of the stacking wall 10, so there is also a gap between the stacking wall 10 and the heat insulating layer 34, and the gap and the pit ventilation structure (not shown)
  • the connection is used for the ventilation airflow passage of the pit during the normal operation of the nuclear power plant core, which is called the normal ventilation flow passage B.
  • the present invention also provides a return flow passage 124 for the external cooling of the stack under the severe accident in the side wall 12 of the pile shield wall 10.
  • the top inlet of the return flow passage 124 is disposed on the bottom wall of the accommodating space 120 outside the support ring 16, the entire passage passes through most of the height of the side wall 12, and the bottom outlet is disposed at a position below the inner wall of the side wall 12 near the bottom of the stack 20. .
  • the top inlet of the return flow passage 124 is located at the lowest point of the bottom wall of the accommodation space 120.
  • a check valve or other similarly actuatable opening stop 126 is provided at the bottom outlet of the return flow passage 124.
  • the blocking member 126 Under normal operating conditions, the blocking member 126 is closed, the stack air supply is not bypassed from the return flow passage 124, and the return flow passage 124 does not affect the normal operation function of the heap pit 20; in severe accident conditions, the blocking member 126 is automatically opened by the passive principle, and the return flow passage 124 is opened to provide a return path for the saturated water, thereby forming a natural circulation flow path for the external cooling with the ascending flow passage A.
  • the vapor-liquid mixed medium is separated into steam and saturated water in the annular accommodating space 120, wherein the steam passes through the main pipe inlet and outlet 122 on the sump shielding wall 10 to discharge the pile pit and take away the heat of the pile pit; the separated saturated water passes through The return flow path 124 flows back to the bottom of the heap pit 20;
  • the rising channel A and the return channel 124 constitute a natural circulation loop for the external cooling:
  • the external cooling medium in the rising channel A is a vapor-liquid mixture, and its density is small; in the reflux channel
  • the external cooling medium in the 124 is saturated water, and the density thereof is large; the difference in density between the rising flow passage A and the return flow passage 124 forms a large natural circulation driving force, so that the cooling fluid flowing through the outer wall of the reactor pressure vessel 30 has The higher flow rate is used to continuously and efficiently cool the reactor pressure vessel 30.
  • the combined structure of the nuclear power plant reactor pressure vessel and the shielding wall of the present invention realizes the use of the difference in density of the cooling medium outside the stack in the rising flow channel A by opening the return flow channel 124 in the side wall 12 of the trench shielding wall 10.
  • a stable natural circulation is formed in the return flow path 124, so that the reactor pressure vessel 30 can be rapidly and continuously cooled under severe accident conditions, thereby effectively retaining the melt in the reactor pressure vessel 30, maintaining the reactor pressure vessel 30. Integrity.

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  • Engineering & Computer Science (AREA)
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  • High Energy & Nuclear Physics (AREA)
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Abstract

一种核电站反应堆压力容器与屏蔽墙的组合结构，其包括堆坑屏蔽墙（10）、由堆坑屏蔽墙（10）围成的堆坑（20）、设置在堆坑（20）内的反应堆压力容器（30）和设置在反应堆压力容器（30）外的保温层（34），保温层（34）与反应堆压力容器（30）之间、保温层（34）与堆坑屏蔽墙（10）之间均存在上下贯通的流道；堆坑屏蔽墙（10）的下部穿设有堆坑注水管线（18）；所述堆坑屏蔽墙（10）的上部内壁开设有容留空间（120），堆坑屏蔽墙（10）中设置有将容留空间（120）与堆坑底部空间连通的回流流道（124）。该组合结构通过在堆坑屏蔽墙中设置回流流道，确保了蒸汽夹带的饱和水能够顺利流回堆坑，因此能够在堆坑内的上升流道和回流流道之间形成稳定自然循环，确保反应堆压力容器在严重事故工况下得到快速持续地冷却。

Description

核电站反应堆压力容器与屏蔽墙的组合结构 技术领域
本发明属于核电站严重事故应对领域，更具体地说，本发明涉及一种能够应对严重事故工况的核电站反应堆压力容器与屏蔽墙的组合结构。
背景技术
自从切尔诺贝利、三哩岛、福岛核电站事故后，世界各国对核电安全的重视达到了前所未有的高度，在对下一代核电站发生严重事故概率提出更高要求的同时，纷纷结合各国在役、在建的核电机组特点提出严重事故缓解策略改进方案。在中国核安全局2014年4月18日发布的新版《核动力厂设计安全规定》(HAF102)和IAEA导则NS-G-1.10中，也要求核电站在设计中应尽实际可能地考虑设置严重事故预防和缓解措施。
压水堆核电站发生严重事故后，如果堆内冷却手段失效，反应堆堆芯将逐渐熔化坍塌而掉入反应堆压力容器底部。此时，如果得不到有效的外部冷却，反应堆压力容器将会被熔穿，导致堆内熔融物滞留失效，堆芯熔融物进入到反应堆堆坑，继而与混凝土底板发生MCCI反应(熔融物与混凝土反应)并释放出危害性的可燃气体，从而使得严重事故状况显著恶化。
为防止堆内熔融物滞留失效，部分已公开压水堆核电站采用堆外冷却方式冷却反应堆压力容器，即：严重事故工况下向堆坑内注水，使反应堆压力容器浸泡在堆坑水体中，通过外壁冷却方式确保反应堆压力容器不被熔穿，从而使得堆内熔融物能够被稳定地滞留在反应堆压力容器内。为防止因临界热流密度不足导致的反应堆压力容器熔穿现象，冷却介质必须以较大的流量流经反应堆压力容器外壁，因此，需要为反应堆压力容器外部冷却介质提供合适的流道设 计，设法使堆外冷却介质产生较强的自然循环效应。一般地，会在反应堆压力容器外壁与其保温层之间设置合适的间隙，并在保温层上下部分别设置出汽口和进水口，严重事故工况下上部出汽口和下部进水口依靠非能动原理自动打开，从而为堆外冷却提供上升流道。由于上部出汽口处排出的汽水混合物中，蒸汽质量份额很小(约1％量级)，主要成分是饱和水，只有设法使这部分饱和水以较小的压降回流到堆坑底部，才能构成强烈稳定的堆外冷却自然循环效应。实验和分析证明，加大堆外冷却的自然循环流量有利于提高临界热流密度，从而可以更好地保证堆外冷却能力。
部分已公开的核电设计中，通过反应堆压力容器支撑环的特殊结构设计为堆外冷却提供自然循环回流流道，但部分核电站反应堆支撑环因抗震设计等需求，其上端面高于堆坑溢流标高，没法由其为堆外冷却提供回流流道。为此，需要调整堆坑结构设计方案，解决其堆外冷却的回流流道问题。此外，堆坑结构紧凑，其功能要求较多，需合理整合正常运行及严重事故下的功能设计，并在此框架下解决堆外冷却的回流流道问题。
发明内容
本发明的目的在于：提供一种核电站反应堆压力容器与屏蔽墙的组合结构，以保证冷却介质在严重事故工况下能够形成稳定自然循环，更有效地对反应堆压力容器进行冷却，确保反应堆压力容器的完整性。
为了实现上述目的，本发明提供了一种核电站反应堆压力容器与屏蔽墙的组合结构，其包括堆坑屏蔽墙、由堆坑屏蔽墙围成的堆坑、设置在堆坑内的反应堆压力容器和设置在反应堆压力容器外的保温层，保温层与反应堆压力容器之间、保温层与堆坑屏蔽墙之间均存在上下贯通的流道；堆坑屏蔽墙的下部穿设有堆坑注水管线；所述堆坑屏蔽墙的上部内壁开设有容留空间，堆坑屏蔽墙中设置有将容留空间与堆坑底部空间连通的回流流道。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述 回流流道包括用于连通堆坑底部空间的底部出口，底部出口处设置有可开启的阻挡件；正常工况下，阻挡件将回流流道的底部出口关闭，避免堆坑正常送风从回流流道旁路；严重事故工况下，阻挡件依靠非能动原理自动打开，使回流流道畅通，确保容留空间中的堆外冷却介质顺利回流到堆坑底部。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述保温层与反应堆压力容器之间的流道为严重事故下堆外冷却介质的上升流道；在严重事故工况下，所述上升流道和回流流道在介质密度差的作用下形成自然循环回路，持续对反应堆压力容器进行外部冷却。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述堆坑屏蔽墙的容留空间通过主管道进出口与堆坑外部空间连通；在严重事故工况下，上升流道中的汽液混合介质在容留空间内分离成蒸汽和饱和水，其中蒸汽通过堆坑屏蔽墙上的主管道进出口排出堆坑并带走堆坑热量，分离出的饱和水通过回流流道重新流回到堆坑底部。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述保温层的底部开设有注水孔，顶部与反应堆压力容器之间留有出汽口；保温层的上部出汽口和下部注水孔平时关闭，在严重事故工况下依靠非能动原理自动打开，从而在反应堆压力容器与保温层之间形成堆外冷却介质的上升流道。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述注水孔全部位于反应堆压力容器的下封头下方。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述回流流道的顶部入口位于堆坑屏蔽墙容留空间的最低点。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述堆坑屏蔽墙的容留空间位于反应堆主管道接管段标高处，且为环形。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述反应堆压力容器的上部连接有多根主管道，主管道一一对应地自主管道进出口 穿出堆坑屏蔽墙；主管道进出口的大小大于主管道的横截面积，从而将容留空间与堆坑外部空间连通。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述环形容留空间的底壁上安装有支撑环，支撑环对所有主管道的根部形成支撑，使得反应堆压力容器悬空收容在堆坑内；回流流道的顶部入口位于支撑环外侧的容留空间底壁上。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述主管道进出口的底壁高于环形容留空间的底壁，回流流道的顶部入口位于环形容留空间底壁的最低点。
作为本发明核电站反应堆压力容器与屏蔽墙的组合结构的一种改进，所述保温层与堆坑屏蔽墙之间的流道为正常通风流道，在正常工况下用做堆坑通风的气流通道。
与现有技术相比，本发明核电站反应堆压力容器与屏蔽墙的组合结构通过在堆坑屏蔽墙中设置回流流道，确保了蒸汽夹带的饱和水能够顺利流回堆坑，因此能够在堆坑内的上升流道和回流流道之间形成稳定自然循环，确保反应堆压力容器在严重事故工况下得到快速持续地冷却，从而有效地将熔融物滞留在反应堆压力容器内。
附图说明
下面结合附图和具体实施方式，对本发明核电站反应堆压力容器与屏蔽墙的组合结构及其有益技术效果进行详细说明，其中：
图1为本发明核电站反应堆压力容器与屏蔽墙的组合结构的示意图。
具体实施方式
为了使本发明的目的、技术方案和有益技术效果更加清晰明白，以下结合附图和具体实施方式，对本发明进行进一步详细说明。应当理解的是，本说明 书中描述的具体实施方式仅仅是为了解释本发明，并不是为了限定本发明。
请参阅图1，本发明核电站反应堆压力容器与屏蔽墙的组合结构包括堆坑屏蔽墙10、由堆坑屏蔽墙10围成的堆坑20和设置在堆坑20内的反应堆压力容器30。反应堆压力容器30在上部连接有自堆坑屏蔽墙10穿出的多根主管道32。
堆坑屏蔽墙10包括侧墙12和底板14。侧墙12在上部的内壁上开有一个环形的容留空间120和多个将容留空间120与堆坑外部空间连通的主管道进出口122；容留空间120位于反应堆主管道接管段的标高处。主管道进出口122的数量和位置与主管道32的数量和位置相对应。容留空间120的底壁上安装有支撑环16，支撑环16对所有主管道32的根部形成支撑，使得反应堆压力容器30悬空收容在堆坑20内。由于主管道进出口122的大小大于主管道32的横截面积，且二者之间未采用密封设置，因此主管道进出口122成为堆坑20及容留空间120与外部空间连通的通道。主管道进出口122的底壁高于容留空间120的底壁，主管道进出口122和容留空间120共同形成堆坑屏蔽墙10的出汽口。侧墙12的下部密封穿设有堆坑注水管线18，堆坑注水管线18的出口位于靠近堆坑20底部的位置，用于在严重事故工况下向堆坑20内注水。堆坑注水管线18与堆坑补水箱、安全壳内换料水箱等水源连接，通过非能动、能动或二者相结合的方式向堆坑20内注水。
为了保证反应堆压力容器30外部冷却的效率，还在其外部设置有保温层34。保温层34通过外部支撑(图未示)与堆坑屏蔽墙10的内壁连接，其内壁与反应堆压力容器30无连接，因此，保温层34的内壁与反应堆压力容器30的外壁之间存在上下贯通的间隙。而且，保温层34的底部开设有注水孔340，注水孔340横向排列在反应堆压力容器30的下封头下方，保温层34的顶部也与反应堆压力容器30之间留有出汽口。保温层34的上部出汽口和下部注水孔340平时关闭，在严重事故工况下则依靠非能动原理自动打开，从而在反应堆压力容器30与保温层34之间形成上升流道A，用于严重事故工况下堆外冷却介质的上升。 保温层34的外壁也并没有与堆坑屏蔽墙10的内壁贴合，因此在堆坑屏蔽墙10与保温层34之间也存在上下贯通的间隙，此间隙与堆坑通风结构(图未示)连接，用于核电站堆芯正常运行期间的堆坑通风气流通道，称为正常通风流道B。
为了使冷却水能够在堆坑20内形成自然循环，本发明还在堆坑屏蔽墙10的侧墙12中设置有用于严重事故下堆外冷却的回流流道124。回流流道124的顶部入口设于支撑环16外侧的容留空间120底壁上，整个通道穿过侧墙12的大部分高度，底部出口设于侧墙12内壁下部靠近堆坑20底部的位置上。为了保证回流顺畅，回流流道124的顶部入口所在位置是容留空间120底壁的最低点。回流流道124的底部出口处设有逆止阀或其他具有类似功能的可开启阻挡件126。在正常工况下，阻挡件126闭合，堆坑送风不会从回流流道124处旁路，回流流道124也不影响堆坑20的正常运行功能；在严重事故工况下，阻挡件126则依靠非能动原理自动打开，回流流道124开启畅通，为饱和水提供回流路径，从而与上升流道A构成堆外冷却的自然循环流道。
使用本发明的核电站在严重事故工况下的工作步骤为：
1)当反应堆压力容器30出口温度达到650℃时，通过堆坑注水管线18向堆坑20内注入冷却水，注水方式为：初始阶段大流量注水以快速充满堆坑20，淹没反应堆压力容器30的底部外壁；后续以小流量注水，以补充因冷却反应堆压力容器而蒸发的水量；
2)保温层34的下部注水孔340和上部出汽口以及设于回流流道124底部出口处的阻挡件126均自动打开；进入堆坑20内的冷却水分为两路：一路通过注水孔340，经由上升流道A对压力容器30的壁面进行冷却；一路进入并留在正常通风流道B中，对堆坑20整体进行冷却；
3)进入上升流道A内的冷却水流经反应堆压力容器30的下封头段时被加热至泡核沸腾，形成汽液混合介质，该混合介质上升到上升流道A的顶部后，经由反应堆压力容器30与保温层34、支撑环16之间的空隙流出，进入容留空 间120；
4)汽液混合介质在环形容留空间120内分离成蒸汽和饱和水，其中蒸汽通过堆坑屏蔽墙10上的主管道进出口122排出堆坑并带走堆坑热量；分离出的饱和水通过回流流道124流回到堆坑20的底部；
5)此时，上升流道A和回流流道124即构成了堆外冷却的自然循环回路：在上升流道A内的堆外冷却介质为汽液混合物，其密度较小；在回流流道124内的堆外冷却介质是饱和水，其密度较大；上升流道A和回流流道124的密度差形成了较大的自然循环驱动力，使得流经反应堆压力容器30外壁的冷却流体具有较高流速，从而持续高效地对反应堆压力容器30进行冷却。
通过以上描述可知，本发明核电站反应堆压力容器与屏蔽墙的组合结构通过在堆坑屏蔽墙10的侧墙12中开设回流流道124，实现了利用堆外冷却介质密度差在上升流道A和回流流道124中形成稳定自然循环，因此在严重事故工况下，能够对反应堆压力容器30进行快速持续地冷却，从而有效地将熔融物滞留在反应堆压力容器30内，保持反应堆压力容器30的完整性。
根据上述原理，本发明还可以对上述实施方式进行适当的变更和修改。因此，本发明并不局限于上面揭示和描述的具体实施方式，对本发明的一些修改和变更也应当落入本发明的权利要求的保护范围内。此外，尽管本说明书中使用了一些特定的术语，但这些术语只是为了方便说明，并不对本发明构成任何限制。

Claims (12)

  1. 一种核电站反应堆压力容器与屏蔽墙的组合结构，包括堆坑屏蔽墙、由堆坑屏蔽墙围成的堆坑、设置在堆坑内的反应堆压力容器和设置在反应堆压力容器外的保温层，保温层与反应堆压力容器之间、保温层与堆坑屏蔽墙之间均存在上下贯通的流道；堆坑屏蔽墙的下部穿设有堆坑注水管线；其特征在于：所述堆坑屏蔽墙的上部内壁开设有容留空间，堆坑屏蔽墙中设置有将容留空间与堆坑底部空间连通的回流流道。
  2. 根据权利要求1所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述回流流道包括用于连通堆坑底部空间的底部出口，底部出口处设置有可开启的阻挡件；正常工况下，阻挡件将回流流道的底部出口关闭，避免堆坑正常送风从回流流道旁路；严重事故工况下，阻挡件依靠非能动原理自动打开，使回流流道畅通，确保容留空间中的堆外冷却介质顺利回流到堆坑底部。
  3. 根据权利要求1或2所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述保温层与反应堆压力容器之间的流道为严重事故下堆外冷却介质的上升流道；在严重事故工况下，所述上升流道和回流流道在介质密度差的作用下形成自然循环回路，持续对反应堆压力容器进行外部冷却。
  4. 根据权利要求3所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述堆坑屏蔽墙的容留空间通过主管道进出口与堆坑外部空间连通；在严重事故工况下，上升流道中的汽液混合介质在容留空间内分离成蒸汽和饱和水，其中蒸汽通过堆坑屏蔽墙上的主管道进出口排出堆坑并带走堆坑热量，分离出的饱和水通过回流流道重新流回到堆坑底部。
  5. 根据权利要求3所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述保温层的底部开设有注水孔，顶部与反应堆压力容器之间留有 出汽口；保温层的上部出汽口和下部注水孔平时关闭，在严重事故工况下依靠非能动原理自动打开，从而在反应堆压力容器与保温层之间形成堆外冷却介质的上升流道。
  6. 根据权利要求5所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述注水孔全部位于反应堆压力容器的下封头下方。
  7. 根据权利要求1所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述回流流道的顶部入口位于堆坑屏蔽墙容留空间的最低点。
  8. 根据权利要求1所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述堆坑屏蔽墙的容留空间位于反应堆主管道接管段标高处，且为环形。
  9. 根据权利要求8所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述反应堆压力容器的上部连接有多根主管道，主管道一一对应地自主管道进出口穿出堆坑屏蔽墙；主管道进出口的大小大于主管道的横截面积，从而将容留空间与堆坑外部空间连通。
  10. 根据权利要求9所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述环形容留空间的底壁上安装有支撑环，支撑环对所有主管道的根部形成支撑，使得反应堆压力容器悬空收容在堆坑内；回流流道的顶部入口位于支撑环外侧的容留空间底壁上。
  11. 根据权利要求9所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述主管道进出口的底壁高于环形容留空间的底壁，回流流道的顶部入口位于环形容留空间底壁的最低点。
  12. 根据权利要求1所述的核电站反应堆压力容器与屏蔽墙的组合结构，其特征在于：所述保温层与堆坑屏蔽墙之间的流道为正常通风流道，在正常工况下用做堆坑通风的气流通道。
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