WO2017028201A1 - 核电站反应堆冷却剂系统主回路的布置结构 - Google Patents

核电站反应堆冷却剂系统主回路的布置结构 Download PDF

Info

Publication number
WO2017028201A1
WO2017028201A1 PCT/CN2015/087352 CN2015087352W WO2017028201A1 WO 2017028201 A1 WO2017028201 A1 WO 2017028201A1 CN 2015087352 W CN2015087352 W CN 2015087352W WO 2017028201 A1 WO2017028201 A1 WO 2017028201A1
Authority
WO
WIPO (PCT)
Prior art keywords
reactor
main
main circuit
nuclear power
pressure vessel
Prior art date
Legal status (The legal status 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 status listed.)
Ceased
Application number
PCT/CN2015/087352
Other languages
English (en)
French (fr)
Inventor
程浩
王庆礼
周媛霞
刘永
张涛
董占发
叶子青
彭国胜
陈兴
李连学
吴春明
曹涛
刘长永
叶镕
赵亮
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
China General Nuclear Power Corp
China Nuclear Power Engineering Co Ltd
Original Assignee
China General Nuclear Power Corp
China Nuclear Power Engineering Co Ltd
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 China General Nuclear Power Corp, China Nuclear Power Engineering Co Ltd filed Critical China General Nuclear Power Corp
Priority to PCT/CN2015/087352 priority Critical patent/WO2017028201A1/zh
Publication of WO2017028201A1 publication Critical patent/WO2017028201A1/zh
Anticipated expiration legal-status Critical
Ceased legal-status Critical Current

Links

Images

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
    • 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
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D1/00—Details of nuclear power plant
    • 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
    • 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 belongs to the field of nuclear power plant design, and more particularly to the arrangement structure of a main circuit of a nuclear power plant reactor coolant system.
  • nuclear power technology which is one of the clean energy technologies, is also constantly developing.
  • the safety of nuclear power is also placed in an increasingly prominent position.
  • the main circuit of the reactor coolant system is arranged in the reactor building.
  • the pressure vessel is located in the center of the reactor building; the three steam generators and the three main pumps are respectively arranged at a 120° angle with the pressure vessel as the center; the voltage regulator is arranged on the heat pipe section of the main pipeline of the first circuit.
  • the height difference L between the center of the main pipe 12 connected to the pressure vessel 10 and the inner surface 14 of the reactor floor of the reactor building is about 12.4 m.
  • the seismic acceleration of the main circuit of the nuclear power plant reactor coolant system is basically designed according to 0.2g
  • the seismic capacity of the reactor fuel assembly is basically designed according to 0.2g.
  • This main circuit has at least the following disadvantages: 1) The seismic acceleration of the reactor is basically designed according to 0.2g, resulting in a smaller range of target sites for nuclear power plants and a relatively small seismic margin; 2) responding to external disasters such as earthquakes In terms of safety, the safety requirements are not high enough; 3) the requirements for the seismic structure design of the reactor fuel assembly are higher and more complex; 4) the internationally recognized three-generation nuclear power technical requirements are not met.
  • the object of the present invention is to provide an arrangement structure of a main circuit of a nuclear power plant reactor coolant system with stronger earthquake resistance to solve the above problems.
  • the present invention provides an arrangement structure of a main circuit of a nuclear power plant reactor coolant system, which comprises a reactor building, a pressure vessel, a steam generator, a voltage regulator, a main pump and a main pipeline, and is provided in the reactor building.
  • a base plate the pressure vessel is installed in the reactor building and located above the base plate;
  • the main pipe is connected between the pressure vessel, the steam generator and the main pump to form a main circuit, and the pressure regulator is arranged on the main pipe;
  • the height difference L between the center of the main pipe connected to the pressure vessel and the inner surface of the reactor floor of the reactor building is 9.7 m to 10.8 m.
  • the height difference L between the center of the main pipe connected to the pressure vessel and the inner surface of the reactor floor of the reactor building is 10.2 m to 10.5 m. .
  • the height difference L between the center of the main pipe connected to the pressure vessel and the inner surface of the reactor floor of the reactor building is 10.2 m to 10.4 m. .
  • the height difference L between the center of the main pipe connected to the pressure vessel and the inner surface of the reactor floor of the reactor building is 10.4 m.
  • the steam generator and the main pump each have three.
  • the three steam generators are arranged symmetrically with each other at an angle of 120° centered on the pressure vessel.
  • the three main pumps are arranged symmetrically at an angle of 120° with each other centered on the pressure vessel.
  • the number of the voltage regulators is one, which is arranged on the heat pipe section of the main pipe.
  • the reactor building is provided with a containment water tank inside, and the structure is a double ring pool structure.
  • the arrangement structure of the main circuit of the nuclear power plant reactor coolant system of the present invention adopts the main circuit sinking arrangement, so that the seismic acceleration of the reactor fuel assembly can be achieved without the large adjustment of the reactor fuel assembly structure design. Increased to 0.3g or more, which improves the seismic capacity of the reactor, enabling it to meet the internationally recognized seismic requirements for third-generation nuclear power technology.
  • FIG. 1 is a schematic view showing the position of a pressure vessel of an arrangement structure of a main circuit of a nuclear power plant reactor coolant system.
  • FIG. 2 is a partial cross-sectional view showing the arrangement structure of a main circuit of a nuclear power plant reactor coolant system according to the present invention.
  • FIG. 3 is a schematic diagram showing the structure of a main circuit of an arrangement structure of a main circuit of a nuclear power plant reactor coolant system according to the present invention.
  • FIG. 4 is a schematic view showing the position of a pressure vessel of an arrangement structure of a main circuit of a nuclear power plant reactor coolant system according to the present invention.
  • the first task is to make the design of the main circuit, especially the reactor fuel assembly, meet the seismic requirements of 0.3 g or more. That is to say, making the seismic capacity of the reactor fuel assembly 0.3 g or more is a necessary condition for the entire power station to achieve a seismic capacity of 0.3 g or more.
  • reactor fuel assemblies are basically unable to improve the seismic capacity by modifying the structural design of the fuel assembly itself, and can only be achieved by other means.
  • the inventors of the present application conducted sensitivity analysis on various main parameters affecting the dynamic response of the plant, repeated trial and error adjustment schemes, and finally found the main pipeline of the main circuit of the reactor coolant system of the nuclear power plant.
  • the height difference L between the center and the surface of the reactor floor inside the reactor building is a critical parameter for the seismic performance of the reactor. The smaller the value of the height difference L, the stronger the seismic performance of the reactor. Therefore, it is conceivable to solve the above problem by reducing the seismic response of the fuel assembly position.
  • the present invention reduces the elevation of the fuel assembly arrangement by the "sinking" arrangement of the reactor, ie the "sinking" arrangement of the main circuit, in order to minimize the central elevation of the main pipeline and the interior of the reactor building.
  • the height difference between the surfaces is improved, and the structural rigidity of the reactor building is improved by optimizing the structural arrangement scheme, and the seismic action of the fuel assembly is reduced to achieve the seismic requirement of 0.3 g or more.
  • the arrangement of the main circuit of the nuclear power plant reactor coolant system of the present invention includes a reactor building, a pressure vessel 30, a steam generator 36, a main pump 38, a main pipe 32, and a voltage regulator (not shown).
  • the reactor building is provided with a crucible base plate, and the pressure vessel 30 is installed in the center of the reactor building and above the crucible base plate.
  • the main pipe 32 is connected between the pressure vessel 30, the steam generator 36, and the main pump 38 to form a main circuit.
  • three steam generators 36 and three main pumps 38 are respectively arranged at a 120° angle with each other centering on the pressure vessel 30; the regulator is disposed on the heat pipe section of the main pipe 32.
  • the refueling tank 200 is provided in the reactor building, and the structure is a double ring pool structure.
  • the height difference L between the center of the main pipe 32 connected to the pressure vessel 30 and the inner surface 34 of the reactor floor of the reactor building is 10.2 m to 10.5 m, preferably 10.4 m. This is because during the research, it was found that the height difference L between the center of the main pipe 32 of the main circuit and the surface of the base plate 34 of the reactor building can meet the seismic requirement of 0.3g of the fuel assembly when the height difference L is between 9.7m and 10.8m. Combined with the layout of the reactor building and the surrounding raft base plant, the height difference L is 10.2m ⁇ 10.5m, which is a relatively reasonable and feasible solution.
  • the choice of height difference of 10.4m is the main circuit "sinking" arrangement
  • the overall coordination of the structure and the overall layout of the reactor building and the surrounding raft base plant layout structure is the best. That is, the optimum height difference L between the center of the main circuit main pipe 32 of the present invention and the inner surface 34 of the reactor building is reduced by 2.0 m from the disclosed nuclear power plant.
  • the present invention also optimizes the related structure: adjusting the wall structure of the main equipment compartment and the partial wall structure of the reactor pool; adjusting the elevation of the cooling and ventilation system of the containment in the reactor building
  • the main equipment load-bearing layer is adjusted to the elevation above the operating platform; the pipeline connected to the main circuit is moved along with the main pipeline 32; the layout scheme of the refueling water tank 200 in the containment is changed from a single-ring pool to a double-ring pool structure.
  • the present invention adopts a main circuit sinking arrangement, and the height difference L between the center of the main circuit main pipe 32 and the inner surface of the reactor floor of the reactor building is less than or equal to 10.4 m, and the height difference is relatively
  • the open nuclear power plant has been reduced by 2.0m or more, so the seismic acceleration of the reactor fuel assembly can be increased to 0.3g or more without making large adjustments in the reactor fuel assembly structure design, thereby improving the seismic capacity of the reactor and enabling it to Reaching the internationally recognized earthquake resistance to the three generations of nuclear power technology begging.
  • the arrangement structure of the main circuit of the nuclear power plant reactor coolant system of the invention has at least the following advantages: 1) using the main circuit sinking arrangement structure, the seismic acceleration of the reactor is increased to 0.3 g or more, and the target site selection will be more in the world. Wide, the seismic margin will be relatively larger, so it can meet the seismic requirements of more sites; 2) greatly enhance the ability of nuclear power plants to cope with external disasters such as earthquakes, and the safety of nuclear power plants has been significantly strengthened, helping to eliminate the public's nuclear power plants.
  • the seismic acceleration of the reactor has increased from 0.2g to 0.3g or more, and the seismic acceleration of nuclear power plants has also increased from 0.2g to 0.3g or more, enabling nuclear power plants to withstand external disasters such as stronger earthquakes and improve public Confidence and acceptance of nuclear power plant safety; 3)
  • the seismic design of nuclear power plants meet the internationally recognized three-generation nuclear power technical requirements, meet the requirements of EUR and URD: the seismic acceleration of the reactor is 0.3g or more, and the three-generation nuclear power technology that has been published Compared with 0.25g and 0.30g, it has reached the internationally recognized technical requirements of three generations of nuclear power users; 4) It can effectively reduce the difficulty in designing the seismic structure of the reactor fuel assembly, avoiding the large design modification of the reactor fuel assembly structure design due to the improvement of seismic requirements, which is beneficial to shorten the fuel assembly structure design time.

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

一种核电站反应堆冷却剂系统主回路的布置结构，其包括反应堆厂房、压力容器（30）、蒸汽发生器（36）、稳压器、主泵（38）和主管道（32），反应堆厂房内设有筏基底板；压力容器（30）安装在反应堆厂房中，并位于筏基底板的上方；主管道（32）连接在压力容器（30）、蒸汽发生器（36）、主泵（38）之间而形成主回路，稳压器布置在主管道（32）上；所述与压力容器（30）连接的主管道（32）的中心和反应堆厂房内部筏基底板上表面之间的高差L为9.7m～10.8m。与现有技术相比，本申请采用了主回路下沉布置，因此能够在反应堆燃料组件结构设计不做大的调整的情况下，将反应堆燃料组件抗震加速度提高到0.3g或以上，从而提高了反应堆抗震能力，使其能够达到国际上公认的对三代核电技术的抗震要求。

Description

核电站反应堆冷却剂系统主回路的布置结构 技术领域
本发明属于核电站设计领域，更具体地说，本发明涉及一种核电站反应堆冷却剂系统主回路的布置结构。
背景技术
随着经济发展和环境保护要求的日益提高，作为清洁能源技术之一的核电技术也在不断向前发展，核电站设计在更新换代过程中，核电的安全性也被摆在愈加突出的位置。
在已公开的核电站中，反应堆冷却剂系统主回路布置在反应堆厂房内。其中，压力容器位于反应堆厂房中心；3台蒸汽发生器和3台主泵分别以压力容器为中心，彼此呈120°夹角对称布置；稳压器布置在一回路主管道的热管段上。请参阅图1，在上述反应堆冷却剂系统的主回路中，与压力容器10连接的主管道12的中心和反应堆厂房内部筏基底板上表面14之间的高差L约为12.4m。
但是，上述核电站反应堆冷却剂系统主回路的抗震加速度基本上是按0.2g进行设计，其反应堆燃料组件的抗震能力也基本上是按0.2g进行设计。这种主回路至少具有以下缺点：1)反应堆的抗震加速度基本上按0.2g进行设计，导致核电站在目标厂址选择上范围更小，抗震裕量也相对较小；2)在应对地震等外部灾害方面，安全性不够高；3)对反应堆燃料组件抗震结构设计方面的要求更高、更复杂；4)达不到国际上公认的三代核电技术要求。
发明内容
本发明的目的在于：提供一种抗震能力更强的核电站反应堆冷却剂系统主回路的布置结构，以解决上述问题。
为了实现上述发明目的，本发明提供了一种核电站反应堆冷却剂系统主回路的布置结构，其包括反应堆厂房、压力容器、蒸汽发生器、稳压器、主泵和主管道，反应堆厂房内设有筏基底板；压力容器安装在反应堆厂房中，并位于筏基底板的上方；主管道连接在压力容器、蒸汽发生器、主泵之间而形成主回路，稳压器布置在主管道上；所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为9.7m～10.8m。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.2m～10.5m。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.2m～10.4m。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.4m。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述蒸汽发生器和主泵各有三台。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述三台蒸汽发生器以压力容器为中心，彼此呈120°夹角对称布置。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述三台主泵以压力容器为中心，彼此呈120°夹角对称布置。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述稳压器的数量为一台，其布置在主管道的热管段上。
作为本发明核电站反应堆冷却剂系统主回路的布置结构的一种改进，所述反应堆厂房内设置有安全壳内换料水箱，其结构为双环池结构。
与现有技术相比，本发明核电站反应堆冷却剂系统主回路的布置结构采用了主回路下沉布置，因此能够在反应堆燃料组件结构设计不做大的调整的情况下，将反应堆燃料组件抗震加速度提高到0.3g或以上，从而提高了反应堆抗震能力，使其能够达到国际上公认的对三代核电技术的抗震要求。
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冷却剂系统主回路的布置结构及其有益效果进行详细说明。
图1为已公开核电站反应堆冷却剂系统主回路的布置结构的压力容器位置示意图。
图2为本发明核电站反应堆冷却剂系统主回路的布置结构的部分剖视示意图。
图3为本发明核电站反应堆冷却剂系统主回路的布置结构的主回路结构示意图。
图4为本发明核电站反应堆冷却剂系统主回路的布置结构的压力容器位置示意图。
具体实施方式
为了使本发明的目的、技术方案及其有益技术效果更加清晰，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，并非为了限定本发明。
为了提高核电站的抗震水平、提高核电站的安全性，满足三代核电技术主要特征参数要求，有必要将核电站抗震设计要求提高至0.3g或以上。为了实现这一目标，首要任务是使主回路特别是反应堆燃料组件的设计满足0.3g或以上抗震要求。也就是说，使反应堆燃料组件的抗震能力达到0.3g或以上是整个电站达到0.3g或以上抗震能力的必要条件。
但是，反应堆燃料组件由于自身结构的特殊性，基本无法通过修改燃料组件本身的结构设计来提高抗震能力，只能通过其他途径实现。为了不对反应堆燃料组件结构设计做大的调整，本申请的发明人对影响厂房动力响应的各个主要参数进行了敏感性分析，反复试算调整方案，终于发现核电站反应堆冷却剂系统主回路的主管道中心和反应堆厂房内部筏基底板上表面之间的高差L对于反应堆抗震性能是至关重要的参数，即高差L的值越小，反应堆的抗震性能就越强。因此，可以考虑通过降低燃料组件位置的地震响应来解决上述问题。
以上述研究为基础，本发明通过反应堆的“下沉”布置，也即主回路“下沉”布置来降低燃料组件布置的标高，以尽可能缩小主管道中心标高与反应堆厂房内部筏基底板上表面之间高差，同时通过优化结构布置方案，提高反应堆厂房内部结构刚度，降低燃料组件承受的地震作用，使之达到0.3g或以上的抗震要求。
请参阅图2和图3，本发明核电站反应堆冷却剂系统主回路的布置结构包括反应堆厂房、压力容器30、蒸汽发生器36、主泵38、主管道32和稳压器(图未示)。反应堆厂房内设有筏基底板，压力容器30安装在反应堆厂房的中心，并位于筏基底板的上方。主管道32连接在压力容器30、蒸汽发生器36、主泵38之间而形成主回路。其中，3台蒸汽发生器36和3台主泵38分别以压力容器30为中心，彼此呈120°夹角对称布置；稳压器布置在主管道32的热管段上。反应堆厂房内设置有安全壳内换料水箱200，其结构为双环池结构。
请参阅图4，与压力容器30连接的主管道32的中心和反应堆厂房内部筏基底板上表面34之间的高差L为10.2m～10.5m，优选为10.4m。这是因为在研究过程中发现，主回路的主管道32中心和反应堆厂房内部筏基底板上表面34之间的高差L在9.7m～10.8m时都能满足燃料组件0.3g抗震要求，而结合反应堆厂房及周边共用筏基厂房的布置情况，高差L在10.2m～10.5m是一个相对合理、可行的方案，进一步综合考虑工程可实施性(比如设备、阀门安装、操作便利 性)、人员通行便利性(相关楼层通道净空要求等)和安全壳内换料水箱200布置方案(外环池施工便利性)等因素，选择高差为10.4m是主回路“下沉”布置结构与反应堆厂房整体布置结构及周边共用筏基厂房布置结构整体协调性最好的方案。也就是说，本发明的主回路主管道32的中心和反应堆厂房内部筏基底板上表面34之间的最佳高差L相对已公开的核电站减少了2.0m。
为了配合主回路“下沉”布置，本发明还同时对相关的结构进行了以下优化：调整主设备隔间墙体结构和反应堆水池局部墙体结构；调整反应堆厂房内安全壳冷却通风系统布置标高，由主设备承重层调至操作平台以上标高布置；使与主回路相连的管道随主管道32一起移动；将安全壳内换料水箱200布置方案由单环池改为双环池结构等。
需要说明的是，虽然主回路主管道32的中心和反应堆厂房内部筏基底板上表面34之间的高差L越小，反应堆的抗震性能就越强，但是L也并不能无限制地缩小，原因如下：一是为了使反应堆厂房内与已公开核电站相比有大致相同条件的布置空间和相似结构，以避免后续出现无法解决的未知问题；二是L过小将会导致外环池202上部的环形空间净高过低，以至于无法满足阀门布置空间要求和人员通行要求，使得反应堆厂房底部安全壳内换料水箱200的外环池202施工困难；三是L过小还会导致周边的安全厂房、燃料厂房局部区域布置不合理，尤其是燃料厂房乏池屏蔽水层厚度将不能满足要求。因此，在利用主回路“下沉”布置来提高反应堆燃料组件抗震能力时，还必须对L的最低值有所限定。
与现有技术相比，本发明采用了主回路下沉布置，其主回路主管道32中心和反应堆厂房内部筏基底板上表面34之间的高差L小于等于10.4m，此高差相对已公开的核电站减少了2.0m或以上，因此能够在反应堆燃料组件结构设计不做大的调整的情况下，将反应堆燃料组件抗震加速度提高到0.3g或以上，从而提高了反应堆抗震能力，使其能够达到国际上公认的对三代核电技术的抗震要 求。
本发明核电站反应堆冷却剂系统主回路的布置结构至少具有以下优点：1)采用主回路下沉布置结构，将反应堆抗震加速度提升至0.3g或以上，在世界范围内目标厂址选择上，范围将更广，抗震裕量也将相对更大，因此可以满足更多厂址的抗震要求；2)极大地提升了核电站应对地震等外部灾害方面能力，核电站安全性得到明显加强，有助于消除公众对核电站安全性疑虑和恐惧心理：反应堆抗震加速度从0.2g提升至0.3g或以上，核电站的抗震加速度也从0.2g提升至0.3g或以上，使核电站能抵御更强的地震等外部灾害，提高公众对核电站安全的信心与接受度；3)在核电站抗震设计方面，达到国际上公认的三代核电技术要求，满足EUR和URD要求：反应堆抗震加速度为0.3g或以上，和目前已公开的三代核电技术的0.25g、0.30g相比，已达到国际公认的三代核电用户技术要求；4)可以有效降低反应堆燃料组件抗震结构设计方面的难度，避免了因为抗震要求的提高而对反应堆燃料组件结构设计进行大的设计修改，有利于缩短燃料组件结构设计时间。
根据上述说明书的揭示和教导，本发明所属领域的技术人员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 (9)

  1. 一种核电站反应堆冷却剂系统主回路的布置结构，包括反应堆厂房、压力容器、蒸汽发生器、稳压器、主泵和主管道，反应堆厂房内设有筏基底板；压力容器安装在反应堆厂房中，并位于筏基底板的上方；主管道连接在压力容器、蒸汽发生器、主泵之间而形成主回路，稳压器布置在主管道上；其特征在于：所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为9.7m～10.8m。
  2. 根据权利要求1所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.2m～10.5m。
  3. 根据权利要求1所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.2m～10.4m。
  4. 根据权利要求1所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述与压力容器连接的主管道的中心和反应堆厂房内部筏基底板上表面之间的高差L为10.4m。
  5. 根据权利要求1至4中任一项所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述蒸汽发生器和主泵各有三台。
  6. 根据权利要求5所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述三台蒸汽发生器以压力容器为中心，彼此呈120°夹角对称布置。
  7. 根据权利要求5所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述三台主泵以压力容器为中心，彼此呈120°夹角对称布置。
  8. 根据权利要求1至4中任一项所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述稳压器的数量为一台，其布置在主管道的热管段 上。
  9. 根据权利要求1至4中任一项所述的核电站反应堆冷却剂系统主回路的布置结构，其特征在于：所述反应堆厂房内设置有安全壳内换料水箱，其结构为双环池结构。
PCT/CN2015/087352 2015-08-18 2015-08-18 核电站反应堆冷却剂系统主回路的布置结构 Ceased WO2017028201A1 (zh)

Priority Applications (1)

Application Number Priority Date Filing Date Title
PCT/CN2015/087352 WO2017028201A1 (zh) 2015-08-18 2015-08-18 核电站反应堆冷却剂系统主回路的布置结构

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
PCT/CN2015/087352 WO2017028201A1 (zh) 2015-08-18 2015-08-18 核电站反应堆冷却剂系统主回路的布置结构

Publications (1)

Publication Number Publication Date
WO2017028201A1 true WO2017028201A1 (zh) 2017-02-23

Family

ID=58050739

Family Applications (1)

Application Number Title Priority Date Filing Date
PCT/CN2015/087352 Ceased WO2017028201A1 (zh) 2015-08-18 2015-08-18 核电站反应堆冷却剂系统主回路的布置结构

Country Status (1)

Country Link
WO (1) WO2017028201A1 (zh)

Cited By (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN106910541A (zh) * 2017-03-30 2017-06-30 上海核工程研究设计院 一种紧凑式核反应堆装换料工艺
CN113569439A (zh) * 2021-06-08 2021-10-29 上海核工程研究设计院有限公司 一种核电厂主回路自动化分析前后处理系统和方法
RU2761866C1 (ru) * 2020-12-30 2021-12-13 Акционерное Общество "Атомэнергопроект" Способ мониторинга системы пассивного отвода тепла из внутреннего объема защитной оболочки и устройство для его осуществления
WO2023078412A1 (zh) * 2021-11-04 2023-05-11 中广核研究院有限公司 反应堆一回路主设备安装方法
RU2806815C1 (ru) * 2022-10-10 2023-11-07 Акционерное общество "Опытное Конструкторское Бюро Машиностроения имени И.И. Африкантова" (АО "ОКБМ Африкантов") Система пассивного отвода тепла из внутреннего объема защитной оболочки водо-водяного энергетического реактора
CN119989103A (zh) * 2025-04-09 2025-05-13 东莞市佛尔盛智能机电股份有限公司 核电站通风系统的风机智能控制方法及系统
CN121257960A (zh) * 2025-09-29 2026-01-02 蚌埠学院 一种利用神经网络的施工质量数据分析方法及系统

Citations (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
EP0013210A1 (fr) * 1978-12-15 1980-07-09 Framatome Installation de réacteur nucléaire avec aménagement d'instrumentation de mesure de flux neutronique
US5610962A (en) * 1995-09-22 1997-03-11 General Electric Company Construction of nuclear power plants on deep rock overlain by weak soil deposits
CN101839467A (zh) * 2010-05-10 2010-09-22 中国核电工程有限公司 一种压水堆核电站蒸汽发生器、主管道和主回路安装方法
CN101916595A (zh) * 2010-07-28 2010-12-15 中国广东核电集团有限公司 一种压水堆核电站反应堆一回路抽真空排气方法
CN102169736A (zh) * 2011-01-19 2011-08-31 中国核工业第五建设有限公司 一种核电站冷却剂系统主管道的安装方法
CN203607107U (zh) * 2013-10-30 2014-05-21 中国广核集团有限公司 用于保证核电站安全的安全系统

Patent Citations (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
EP0013210A1 (fr) * 1978-12-15 1980-07-09 Framatome Installation de réacteur nucléaire avec aménagement d'instrumentation de mesure de flux neutronique
US5610962A (en) * 1995-09-22 1997-03-11 General Electric Company Construction of nuclear power plants on deep rock overlain by weak soil deposits
CN101839467A (zh) * 2010-05-10 2010-09-22 中国核电工程有限公司 一种压水堆核电站蒸汽发生器、主管道和主回路安装方法
CN101916595A (zh) * 2010-07-28 2010-12-15 中国广东核电集团有限公司 一种压水堆核电站反应堆一回路抽真空排气方法
CN102169736A (zh) * 2011-01-19 2011-08-31 中国核工业第五建设有限公司 一种核电站冷却剂系统主管道的安装方法
CN203607107U (zh) * 2013-10-30 2014-05-21 中国广核集团有限公司 用于保证核电站安全的安全系统

Cited By (9)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN106910541A (zh) * 2017-03-30 2017-06-30 上海核工程研究设计院 一种紧凑式核反应堆装换料工艺
RU2761866C1 (ru) * 2020-12-30 2021-12-13 Акционерное Общество "Атомэнергопроект" Способ мониторинга системы пассивного отвода тепла из внутреннего объема защитной оболочки и устройство для его осуществления
WO2022146189A1 (ru) * 2020-12-30 2022-07-07 Акционерное Общество "Атомэнергопроект" Способ и устройство мониторинга системы пассивного отвода тепла
CN113569439A (zh) * 2021-06-08 2021-10-29 上海核工程研究设计院有限公司 一种核电厂主回路自动化分析前后处理系统和方法
WO2023078412A1 (zh) * 2021-11-04 2023-05-11 中广核研究院有限公司 反应堆一回路主设备安装方法
RU2806815C1 (ru) * 2022-10-10 2023-11-07 Акционерное общество "Опытное Конструкторское Бюро Машиностроения имени И.И. Африкантова" (АО "ОКБМ Африкантов") Система пассивного отвода тепла из внутреннего объема защитной оболочки водо-водяного энергетического реактора
CN119989103A (zh) * 2025-04-09 2025-05-13 东莞市佛尔盛智能机电股份有限公司 核电站通风系统的风机智能控制方法及系统
CN119989103B (zh) * 2025-04-09 2025-08-12 东莞市佛尔盛智能机电股份有限公司 核电站通风系统的风机智能控制方法及系统
CN121257960A (zh) * 2025-09-29 2026-01-02 蚌埠学院 一种利用神经网络的施工质量数据分析方法及系统

Similar Documents

Publication Publication Date Title
WO2017028201A1 (zh) 核电站反应堆冷却剂系统主回路的布置结构
CN104167226B (zh) 一种可实现临界及次临界运行实验的液态金属冷却反应堆实验系统
CN202855316U (zh) 压水堆核电厂安全壳冷却系统
Reyes Jr NuScale plant safety in response to extreme events
RU2596160C2 (ru) Небольшая атомная электростанция на быстрых нейтронах с длительным интервалом замены топлива
CN205104238U (zh) 内置式蒸汽稳压器
CN103850483A (zh) 一种核电厂主厂房群布置方法
CN107564592B (zh) 一种非能动风冷却安全壳
CN104637553A (zh) 流量分配装置及具有该装置的核反应堆组件
CN102637465A (zh) 一种非能动安全壳冷却系统
CN103065697A (zh) 一种不置信硼的乏燃料水池格架临界安全控制方法
CN104376880A (zh) 钢制安全壳能量控制系统
CN106469578B (zh) 核电站反应堆冷却剂系统主回路的布置结构
CN106098121B (zh) 气液流通管及安全壳
CN202126850U (zh) 蓄水型非能动双层安全壳
CN201126717Y (zh) 池式钠冷快堆事故余热排放系统
CN205354669U (zh) 一种具有屏蔽功能的反应堆压力容器支承结构
GB2533829A (en) Active-combined passive containment cooling system
CN106205747A (zh) 池式供热反应堆及低温核供热系统
CN110689985B (zh) 一种托卡马克磁约束聚变电站主厂房群的布置方法及结构
CN203366752U (zh) 非能动压水堆降压系统
US20180301234A1 (en) Residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant
CN109920563B (zh) 一种适用于耐高温驱动机构的桁架式一体化堆顶结构
CN113808768A (zh) 一种双堆共用反应堆厂房的核能供热堆布置方法
Brown et al. PPPL ST-FNSF engineering design details

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: 15901405

Country of ref document: EP

Kind code of ref document: A1

NENP Non-entry into the national phase

Ref country code: DE

122 Ep: pct application non-entry in european phase

Ref document number: 15901405

Country of ref document: EP

Kind code of ref document: A1

122 Ep: pct application non-entry in european phase

Ref document number: 15901405

Country of ref document: EP

Kind code of ref document: A1