JPH02222876A - Emergency core cooling system - Google Patents
Emergency core cooling systemInfo
- Publication number
- JPH02222876A JPH02222876A JP1017285A JP1728589A JPH02222876A JP H02222876 A JPH02222876 A JP H02222876A JP 1017285 A JP1017285 A JP 1017285A JP 1728589 A JP1728589 A JP 1728589A JP H02222876 A JPH02222876 A JP H02222876A
- Authority
- JP
- Japan
- Prior art keywords
- reactor
- coolant
- pressure
- vessel
- tank
- 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.)
- Pending
Links
- 238000001816 cooling Methods 0.000 title claims description 28
- 239000002826 coolant Substances 0.000 claims abstract description 30
- 238000002347 injection Methods 0.000 claims abstract description 11
- 239000007924 injection Substances 0.000 claims abstract description 11
- 230000007246 mechanism Effects 0.000 claims abstract description 7
- 230000007423 decrease Effects 0.000 claims description 6
- 239000000498 cooling water Substances 0.000 abstract description 13
- 230000015556 catabolic process Effects 0.000 abstract 3
- 238000001514 detection method Methods 0.000 abstract 1
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 14
- 238000010586 diagram Methods 0.000 description 2
- 230000000712 assembly Effects 0.000 description 1
- 238000000429 assembly Methods 0.000 description 1
- 230000000052 comparative effect Effects 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
Classifications
-
- 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
Landscapes
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
【発明の詳細な説明】 ［発明の目的］ （産業上の利用分野） 本発明は、原子炉の非常用炉心冷却系に関す−る。[Detailed description of the invention] [Purpose of the invention] (Industrial application field) The present invention relates to an emergency core cooling system for a nuclear reactor.
（従来の技術）
一般に、原子力発電所等の原子炉には、非常用炉心冷却
系（ＥＣＣ３）が配設されており、原子炉圧力容器バウ
ンダリの破断が発生し、冷却材喪失事故に至った場合、
この非常用炉心冷却系により炉心冷却を行い、健全性を
確保することができるよう構成されている。(Prior art) Generally, emergency core cooling systems (ECC3) are installed in nuclear reactors such as nuclear power plants, and rupture of the reactor pressure vessel boundary occurs, leading to a loss of coolant accident. case,
The emergency core cooling system is configured to cool the core and ensure its integrity.
従来、このような非常用炉檀１冷却系は、動的機器であ
るポンプを用いて炉心に冷却材（冷却水）を送るよう構
成されている。また、一般に非常用炉心冷却系は、高圧
非常用炉心冷却系と低圧非常用炉心冷却系とから構成さ
れており、各種の破断事故に対しても適切に対処するこ
とができるよう構成されている。Conventionally, such an emergency reactor 1 cooling system is configured to send coolant (cooling water) to the reactor core using a pump, which is a dynamic device. In addition, the emergency core cooling system generally consists of a high-pressure emergency core cooling system and a low-pressure emergency core cooling system, and is configured to be able to appropriately respond to various types of breakage accidents. .
すなわち、たとえば小口径の破断が生じた場合は、破断
口からの冷却材流失流量が比較的少ないため、原子炉圧
力が高圧に保たれている。そこで、このような場合は高
圧非常用炉心冷却系を用いて原子炉内に冷却材を注入し
、炉心を露出させることなく冷却する。That is, when a small-diameter fracture occurs, for example, the amount of coolant flowing out from the fracture port is relatively small, so that the reactor pressure is maintained at a high pressure. Therefore, in such a case, a high-pressure emergency core cooling system is used to inject coolant into the reactor to cool the reactor without exposing it.
ま・た、たとえば大口径の破断が生じた場合は、破断口
からの冷却材流失流量が多く炉心は一時的に露出し、原
子炉圧力も大幅に低下する。そこで、このような場合は
大容量の低圧ｌｌ；常用炉心冷却系を用いて原子炉内に
冷却材を注入し炉心を再冠水させて冷却する。For example, if a large-diameter rupture occurs, a large amount of coolant flows out from the rupture port, and the reactor core is temporarily exposed, resulting in a significant drop in reactor pressure. Therefore, in such a case, a large-capacity, low-pressure normal core cooling system is used to inject coolant into the reactor and re-flood the core to cool it.
一方、近年非常用炉心冷却系の信頼性向上の観点から、
様々な非常用炉心冷却系が提案されおり−１たとえば加
圧水型原子力発電所では、高圧タンクを用いた非常用炉
心冷却系が採用されている。すなわち、原子炉圧力容器
に冷却材注入配管を介して冷却材を収容した高圧タンク
を接続し、この冷却材注入配管に、原子炉圧力容器内の
圧力が所定の圧力以下に低下すると開となる弁機構（逆
止弁）を設けることにより、冷却材喪失事故が発生し、
原子炉圧力容器内の圧力が低下すると自動的に高圧タン
ク内の冷却材が原子炉圧力容器内に流入するよう構成し
たものである。On the other hand, in recent years, from the perspective of improving the reliability of the emergency core cooling system,
Various emergency core cooling systems have been proposed. For example, in pressurized water nuclear power plants, emergency core cooling systems using high pressure tanks are employed. In other words, a high-pressure tank containing coolant is connected to the reactor pressure vessel via a coolant injection pipe, and the coolant injection pipe opens when the pressure inside the reactor pressure vessel drops below a predetermined pressure. By installing a valve mechanism (check valve), coolant loss accidents can occur.
The structure is such that when the pressure inside the reactor pressure vessel decreases, the coolant in the high pressure tank automatically flows into the reactor pressure vessel.
（発明が解決しようとする課題）
しかしながら、上述したような非常用炉心冷却系におい
ても、さらに効率的な炉心の冷却を可能とし、原子炉の
安全性を向上させることが望まれている。(Problems to be Solved by the Invention) However, even in the emergency core cooling system as described above, it is desired to enable more efficient cooling of the core and improve the safety of the nuclear reactor.
本発明はかかる従来の事情に対処してなされたもので、
従来に較べて効率的に炉心の冷却を行うことができ、原
子炉の安全性の向上を図ることのできる非常用炉心冷却
系を提供しようとするものである。The present invention has been made in response to such conventional circumstances,
The present invention aims to provide an emergency core cooling system that can cool the reactor core more efficiently than in the past and improve the safety of the reactor.
［発明の構成］
（課題を解決するための手段）
すなわち、本発明の非常用炉心冷却系は、冷却材注入配
管を介して原子炉圧力容器に接続され内部に収容した冷
却材を前記原子炉圧力容器内に注入可能に加圧状態とさ
れた複数のタンクと、前記冷却材の前記原子炉圧力容器
内への流入を制御するための弁機構と、前記複数のタン
クのうち先に前記原子炉圧力容器内に前記冷却材の注入
を開始したタンク内の冷却材量の減少を検知して池のタ
ンクの前記弁機構を操作し、これらのタンク内の前記冷
却材が前記原子炉圧力容器内に連続的に注入されるよう
制御する制御装置とを備えたことを特徴とする。[Structure of the Invention] (Means for Solving the Problems) That is, the emergency core cooling system of the present invention is connected to a reactor pressure vessel via a coolant injection pipe, and supplies the coolant stored therein to the reactor. a plurality of tanks pressurized so as to be injectable into the pressure vessel; a valve mechanism for controlling the flow of the coolant into the reactor pressure vessel; and a valve mechanism for controlling the flow of the coolant into the reactor pressure vessel; The valve mechanism of the pond tank is operated by detecting a decrease in the amount of coolant in the tanks that have started injection of the coolant into the reactor pressure vessel, and the coolant in these tanks is injected into the reactor pressure vessel. The invention is characterized by comprising a control device for controlling continuous injection into the container.
（作 用）
上記構成の本発明の非常用炉心冷却系では、従来に較べ
て効率的に炉心の冷却を行うことができ、原子炉の安全
性の向上を図ることができる。(Function) In the emergency core cooling system of the present invention having the above configuration, the core can be cooled more efficiently than in the past, and the safety of the nuclear reactor can be improved.
（実施例）
以下、本発明の詳細を図面を参照して一実施例について
説明する。(Example) Hereinafter, details of the present invention will be described with reference to the drawings.
第１図は、本発明の一実施例の非常用炉心冷却系の構成
を示すもので、図において符号１は、原子炉圧力容器を
示している。この原子炉圧力容器１内には多数の燃料集
合体（図示せず）から構成された炉心２が配設されてい
る。FIG. 1 shows the configuration of an emergency core cooling system according to an embodiment of the present invention, and in the figure, reference numeral 1 indicates a reactor pressure vessel. A reactor core 2 made up of a large number of fuel assemblies (not shown) is disposed within the reactor pressure vessel 1 .
上記原子炉圧力容器１の上方には高圧タンク３と中圧タ
ンク４が配置されており、これらの高圧タンク３および
中圧タンク４は、それぞれ冷却材注入配管５．６を介し
て原子炉圧力容器１に接続されている。A high-pressure tank 3 and an intermediate-pressure tank 4 are arranged above the reactor pressure vessel 1, and these high-pressure tank 3 and intermediate-pressure tank 4 are connected to the reactor pressure via coolant injection piping 5.6, respectively. Connected to container 1.
また、高圧タンク３に接続された冷却材注入配管５には
、通常時は閉とされ、原子炉圧力容器１内の水位低下信
号（図示せず）により開とされる逆止弁７が介挿されて
おり、高圧タンク３には、水位センサ８が設けられてい
る。In addition, a check valve 7 is interposed in the coolant injection pipe 5 connected to the high-pressure tank 3, which is normally closed and is opened by a water level drop signal (not shown) in the reactor pressure vessel 1. A water level sensor 8 is provided in the high pressure tank 3.
さらに、中圧タンク４に接続された冷却材注入配管５に
は、通常時は閉とされ、上記逆止弁７の開信号７ａおよ
び水位センサ８による高圧タンク３内の水位低下信号８
ａにより制御装置９によって開とされる逆止弁１０が介
挿されている。Further, the coolant injection pipe 5 connected to the medium pressure tank 4 is normally closed, and includes an open signal 7a of the check valve 7 and a water level drop signal 8 in the high pressure tank 3 from the water level sensor 8.
A check valve 10 that is opened by the control device 9 is inserted.
上記構成のこの実施例の非常用炉心冷却系では、大口径
の破断が生じた場合は、原子炉圧力容器１内の水位低下
信号により、まず逆止弁７が開となり、高圧タンク３内
の冷却水が原子炉圧力容器］。In the emergency core cooling system of this embodiment with the above configuration, when a large diameter rupture occurs, the check valve 7 is first opened by the water level drop signal in the reactor pressure vessel 1, and the water level in the high pressure tank 3 is opened. cooling water is the reactor pressure vessel].
内に流入し、炉心２を冷却する。そして、高圧タンク３
内の水位が低下して水位センサ８から水位低下信号８ａ
が発せられると、この水位低下信号８ａと逆止弁７の開
信号７ａとを入力された制御装置９によって逆止弁１０
が開とされ、中圧タンク４内の冷却水が原子炉圧力容器
１内に流入し、炉心２を冷却する。and cools the reactor core 2. And high pressure tank 3
When the water level in
When the water level drop signal 8a and the open signal 7a of the check valve 7 are input, the control device 9 activates the check valve 10.
is opened, and the cooling water in the intermediate pressure tank 4 flows into the reactor pressure vessel 1 to cool the reactor core 2.
また、小口径の破断が生じた場合は、原子炉圧力容器１
内の圧力が高圧に維持され、かつ原子炉水位は低下する
。この水位低の信号により、自動減圧装置が作動すると
ともに逆止弁７が開となる。In addition, if a small diameter rupture occurs, the reactor pressure vessel 1
The pressure inside the reactor is maintained at high pressure, and the reactor water level is lowered. This low water level signal activates the automatic pressure reducing device and opens the check valve 7.
この後、大口径の破断の場合と同様に原子炉圧力容器１
内の圧力が低下し、高圧タンク３内の水位低下信号８ａ
と逆止弁７の開信号７ａとにより逆止弁１０が開とされ
る。After this, as in the case of a large-diameter rupture, the reactor pressure vessel 1
The pressure inside the high pressure tank 3 decreases, and the water level decrease signal 8a inside the high pressure tank 3
The check valve 10 is opened by the open signal 7a of the check valve 7.
すなわち、大口径の破断が生じた場合も小口径の破断が
生じた場合も同様に、縦軸を原子炉圧力容器１内に流入
する冷却水の流量、横軸を時間とした第２図のグラフに
示すように、まず高圧タンク３内の冷却水が原子炉圧力
容器１内に流入しく曲線Ａ）、この後間断なく連続して
中圧タンク４内の冷却水が原子炉圧力容器１内に流入す
る（曲線Ｂ）。In other words, whether a large-diameter rupture occurs or a small-diameter rupture occurs, the vertical axis is the flow rate of the cooling water flowing into the reactor pressure vessel 1, and the horizontal axis is the time in Fig. 2. As shown in the graph, the cooling water in the high pressure tank 3 first flows into the reactor pressure vessel 1 (curve A), and then the cooling water in the intermediate pressure tank 4 flows into the reactor pressure vessel 1 without interruption. (curve B).
したがって、たとえば第３図のグラフに示すように、原
子炉圧力容器１内の圧力低下に伴って逆止弁７が開とな
った直後に逆止弁１０も開となり、高圧タンク３内の冷
却水の流入（曲線ａ）と、中圧タンク４内の冷却水の流
入（曲線ｂ）とが重複して行われるようなことはない。Therefore, as shown in the graph of FIG. 3, for example, immediately after the check valve 7 opens as the pressure in the reactor pressure vessel 1 decreases, the check valve 10 also opens, and the cooling inside the high pressure tank 3 is reduced. The inflow of water (curve a) and the inflow of cooling water in the intermediate pressure tank 4 (curve b) are not performed at the same time.
なお、このようなｆｆｉ複があると、必要量以上の冷却
水が原子炉圧力容器１内へ流入し、炉心２の冷却にほと
んど寄与することなく破断口を通って原子炉圧力容器１
外へ流失してしまうことになり、その結果実施例の非常
用炉心冷却系の場合に較べて短時間しか炉心２の冷却を
行うことができなくなる。In addition, when such ffi duplication occurs, more cooling water than the required amount flows into the reactor pressure vessel 1 and flows through the rupture port into the reactor pressure vessel 1 without contributing much to the cooling of the reactor core 2.
As a result, the core 2 can only be cooled for a shorter time than in the case of the emergency core cooling system of the embodiment.
すなわち、この実施例の非常用炉心冷却系では、高圧タ
ンク３内の冷却水と中圧タンク４内の冷却水を、間断な
くかつ重複することなく、原子炉圧力容器２内に流入さ
せることが可能となる。したがって、効率的に、より長
時間にわたって炉心２の冷却を行うことができる。That is, in the emergency core cooling system of this embodiment, the cooling water in the high pressure tank 3 and the cooling water in the intermediate pressure tank 4 can flow into the reactor pressure vessel 2 without interruption and without overlap. It becomes possible. Therefore, the core 2 can be efficiently cooled for a longer period of time.
［発明の効果］
以上説明したように、本発明の非常用炉心冷却系によれ
ば、従来に較べて効率的に炉心の冷却を行うことができ
、原子炉の安全性の向上を図ることができる。[Effects of the Invention] As explained above, according to the emergency core cooling system of the present invention, the core can be cooled more efficiently than in the past, and the safety of the reactor can be improved. can.
第１図は本発明の一実施例の非常用炉心冷却系の構成を
示す図、第２図は第１図に示す非常用炉心冷却系におけ
る高圧タンクおよび中圧タンクから原子炉圧力容器内へ
流入する冷却水流量の時間変化を示すグラフ、第３図は
比較例における高圧タンクおよび中圧タンクから原子炉
圧力容器内へ流入する冷却水流量の時間変化を示すグラ
フである。
１・・・・・・・・・・・・原子炉圧力容器２・・・・
・・・・・・・・炉心
３・・・・・・・・・・・・高圧タンク４・・・・・・
・・・・・・中圧タンク５．６・・・・・・冷却材注入
配管
７．１０・・・逆止弁
８・・・・・・・・・・・・水位センサ９・・・・・・
・・・・・・制御装置
出願人 日本原子力事業株式会社出願人
株式会社 東芝
代理人 弁理士 須 山 佐 −
第１（！ｉFigure 1 is a diagram showing the configuration of an emergency core cooling system according to an embodiment of the present invention, and Figure 2 is a diagram showing the structure of the emergency core cooling system shown in Figure 1, from the high pressure tank and intermediate pressure tank to the inside of the reactor pressure vessel. FIG. 3 is a graph showing the time change in the flow rate of cooling water flowing into the reactor pressure vessel from the high pressure tank and the intermediate pressure tank in a comparative example. 1......Reactor pressure vessel 2...
・・・・・・・・・Core 3・・・・・・・・・High pressure tank 4・・・・・・
...... Medium pressure tank 5.6 ... Coolant injection pipe 7.10 ... Check valve 8 ... Water level sensor 9 ... ...
...Control device applicant Japan Atomic Energy Corporation applicant
Toshiba Corporation Representative Patent Attorney Satoshi Suyama - 1st (!i
Claims (1)
れ内部に収容した冷却材を前記原子炉圧力容器内に注入
可能に加圧状態とされた複数のタンクと、前記冷却材の
前記原子炉圧力容器内への流入を制御するための弁機構
と、前記複数のタンクのうち先に前記原子炉圧力容器内
に前記冷却材の注入を開始したタンク内の冷却材量の減
少を検知して他のタンクの前記弁機構を操作し、これら
のタンク内の前記冷却材が前記原子炉圧力容器内に連続
的に注入されるよう制御する制御装置とを備えたことを
特徴とする非常用炉心冷却系。(1) A plurality of tanks connected to the reactor pressure vessel via coolant injection piping and pressurized so that the coolant contained therein can be injected into the reactor pressure vessel; A valve mechanism for controlling the flow into the reactor pressure vessel, and detecting a decrease in the amount of coolant in the tank that started injecting the coolant into the reactor pressure vessel first among the plurality of tanks. and a control device for controlling the valve mechanisms of other tanks so that the coolant in these tanks is continuously injected into the reactor pressure vessel. Core cooling system.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1017285A JPH02222876A (en) | 1989-01-26 | 1989-01-26 | Emergency core cooling system |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1017285A JPH02222876A (en) | 1989-01-26 | 1989-01-26 | Emergency core cooling system |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH02222876A true JPH02222876A (en) | 1990-09-05 |
Family
ID=11939712
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP1017285A Pending JPH02222876A (en) | 1989-01-26 | 1989-01-26 | Emergency core cooling system |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH02222876A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2008020234A (en) * | 2006-07-11 | 2008-01-31 | Hitachi Ltd | Emergency core cooling system |
-
1989
- 1989-01-26 JP JP1017285A patent/JPH02222876A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2008020234A (en) * | 2006-07-11 | 2008-01-31 | Hitachi Ltd | Emergency core cooling system |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US5377242A (en) | Method and system for emergency core cooling | |
| US7983376B2 (en) | Boiling water nuclear reactor and emergency core cooling system of the same | |
| KR101447028B1 (en) | Multi stage safety injection device and passive safety injection system having the same | |
| JPH0715507B2 (en) | Passive safety equipment for nuclear reactors | |
| JPH0666985A (en) | Method for reducing leakage from heat-transfer pipe of pressurized water reactor and steam generator | |
| KR20140023067A (en) | Separate type safety injection tank | |
| JPH0411836B2 (en) | ||
| JPH05307094A (en) | Reactor cooling system of boiling water type nuclear reactor | |
| EP0512761A1 (en) | Standby safety injection system for nuclear reactor plants | |
| JPH02222876A (en) | Emergency core cooling system | |
| EP0585499B1 (en) | Low pressure coolant injection modification for boiling water reactors | |
| US4064001A (en) | Hot leg relief system | |
| JPH05142380A (en) | Emergency core cooling system | |
| WO2021138806A1 (en) | Safety system for handling severe accident of nuclear power plant and control method therefor | |
| JPH02243996A (en) | Reactor having integral type pressure vessel construction | |
| JP2823984B2 (en) | Containment vessel cooling system | |
| JPH05215886A (en) | Emergency reactor core cooling system | |
| JPH03216592A (en) | Cooling water supplying device of boiling water nuclear reactor | |
| JPH0283495A (en) | Core cooling apparatus for emergency | |
| JPS6314794B2 (en) | ||
| JPS62228197A (en) | Light water type reactor | |
| JPH0660947B2 (en) | Suppressor for water level rise of moderator in control rod guide tube | |
| JPH01197695A (en) | Method for preventing water hammer on control rod drive system device | |
| JPS62100696A (en) | Method of automatically decompressing nuclear reactor | |
| JPS5918486A (en) | Control rod drive hydraulic device |