JPH02253194A - Nuclear reactor system - Google Patents
Nuclear reactor systemInfo
- Publication number
- JPH02253194A JPH02253194A JP1073925A JP7392589A JPH02253194A JP H02253194 A JPH02253194 A JP H02253194A JP 1073925 A JP1073925 A JP 1073925A JP 7392589 A JP7392589 A JP 7392589A JP H02253194 A JPH02253194 A JP H02253194A
- Authority
- JP
- Japan
- Prior art keywords
- steam
- pressure
- water
- pump
- reactor
- 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.)
- Granted
Links
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims abstract description 62
- 238000002347 injection Methods 0.000 claims description 17
- 239000007924 injection Substances 0.000 claims description 17
- 239000012530 fluid Substances 0.000 claims description 3
- 238000010795 Steam Flooding Methods 0.000 claims description 2
- 239000000498 cooling water Substances 0.000 abstract description 11
- 238000001816 cooling Methods 0.000 abstract description 2
- 230000003134 recirculating effect Effects 0.000 abstract 3
- 239000000463 material Substances 0.000 abstract 1
- 239000002826 coolant Substances 0.000 description 4
- 238000010586 diagram Methods 0.000 description 4
- 238000009835 boiling Methods 0.000 description 2
- 238000012423 maintenance Methods 0.000 description 2
- 238000010248 power generation Methods 0.000 description 2
- 238000009833 condensation Methods 0.000 description 1
- 230000005494 condensation Effects 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 230000007774 longterm Effects 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)
- Jet Pumps And Other Pumps (AREA)
Abstract
Description
【発明の詳細な説明】
〔発明の目的〕
（産業上の利用分野）
本発明は沸騰水型原子炉の原子炉圧力容器内への注水手
段を改良した原子炉システムに関する。DETAILED DESCRIPTION OF THE INVENTION [Object of the Invention] (Industrial Application Field) The present invention relates to a nuclear reactor system in which a means for injecting water into a reactor pressure vessel of a boiling water nuclear reactor is improved.
（従来の技術）
沸騰水型原子炉では炉心の冷却に普通の水（軽水を用い
ている。すなわち、第４図に示したようなシステムが採
られている。第４図において、炉心１を内蔵する原子炉
圧力容器２には主蒸気管３が接続されている。炉心１で
発生した熱を水で冷却して蒸気とし、その蒸気は主蒸気
管３を通して図示してないタービンへ供給し発電に供し
ている。(Prior art) In a boiling water reactor, ordinary water (light water) is used to cool the reactor core. In other words, a system as shown in Fig. 4 is adopted. In Fig. 4, the reactor core 1 is A main steam pipe 3 is connected to the built-in reactor pressure vessel 2.The heat generated in the reactor core 1 is cooled with water and turned into steam, and the steam is supplied to a turbine (not shown) through the main steam pipe 3. It is used for power generation.
炉心１への冷却水の循環には大形の再循環ポンプ４によ
って再循環系配管７から原子炉圧力容器２内へ供給して
いる。また、原子炉の運転を停止する場合には炉心１の
余剰発熱を除去する熱除去ポンプ５を駆動させ熱交換器
６で冷却させた低温水を原子炉圧力容器２内へ供給する
。この熱除去ポンプ５は非常用タンク１０から導出され
た熱除去系配管５ａによる給水と再循環系配管７からの
給水とによって原子炉圧力容器２内へ水を供給している
。To circulate cooling water to the reactor core 1, it is supplied into the reactor pressure vessel 2 from a recirculation system piping 7 by a large recirculation pump 4. Further, when the operation of the nuclear reactor is to be stopped, the heat removal pump 5 for removing excess heat from the reactor core 1 is driven to supply low-temperature water cooled by the heat exchanger 6 into the reactor pressure vessel 2. The heat removal pump 5 supplies water into the reactor pressure vessel 2 using water supplied from a heat removal system piping 5a led out from an emergency tank 10 and water supplied from a recirculation system piping 7.
さらに、再循環系配管７の仮想事故破断などに対応して
は原子炉を急緊に停止するとともに迅速にかつ長期間の
冷却が必要である。そのため高圧時に作動させる高圧注
入系配管８ａを設けるとともに低圧時に作動させる低圧
注入系配管９ａをも設けている。これらの系統配管８ａ
、９ａは非常用タンク１０から各ポンプ８，９を介して
原子炉圧力容器２に接続している。これらのポンプ８，
９を作動させて原子炉圧力容器２内の炉心１へ注入して
いる。Furthermore, in response to a hypothetical accidental rupture of the recirculation system piping 7, it is necessary to urgently shut down the nuclear reactor and to perform rapid and long-term cooling. For this reason, a high-pressure injection system piping 8a that is activated when the pressure is high is provided, and a low-pressure injection system piping 9a that is activated when the pressure is low is also provided. These system piping 8a
, 9a are connected from the emergency tank 10 to the reactor pressure vessel 2 via respective pumps 8 and 9. These pumps 8,
9 is activated to inject into the reactor core 1 inside the reactor pressure vessel 2.
（発明が解決しようとする課題）
」二連したように炉心１の冷却には軽水を使用し、発電
後の復水または炉水位の調整の通常運転時の原子炉圧力
容器２内への注入にはモータで駆動するポンプを使用し
ている。また原子炉の通常の停止時には炉心１の残存熱
の除去が必要であり、この場合にもモータ駆動のポンプ
を使用して注水をしている。とくに、仮想する原子炉の
事故において、たとえば配管破断による冷却材喪失事故
時の原子炉圧力容器内への緊急給水時には、従来はモー
タ駆動のポンプを使用したり、電源の不具合を考慮ジー
ゼルエンジンを使用したポンプによっている。(Problem to be solved by the invention) As described above, light water is used to cool the reactor core 1, and light water is injected into the reactor pressure vessel 2 during normal operation for condensation after power generation or for adjusting the reactor water level. uses a pump driven by a motor. Further, during normal shutdown of the nuclear reactor, it is necessary to remove residual heat in the reactor core 1, and in this case as well, water is injected using a motor-driven pump. In particular, in the case of a hypothetical nuclear reactor accident, for example, when emergency water is supplied to the reactor pressure vessel in the event of loss of coolant due to pipe rupture, conventionally motor-driven pumps have been used, or diesel engines have been used in consideration of power supply failures. It depends on the pump used.
しかしながら、緊急時にしか作動しない機器の場合では
通常時のメンテナンスが負担となり、また常用として使
用する機器の場合も含めて運転の動力が全体の効率を考
慮すると損失になる課題がある。However, in the case of equipment that operates only in emergencies, maintenance during normal times becomes a burden, and even in the case of equipment that is used regularly, there is a problem in that the operating power is lost when considering overall efficiency.
本発明は上記課題を解決するためになされたもので、構
造上駆動部のない、たとえば原子炉圧力容器から排出さ
れる蒸気を使用して、蒸気の潜熱を利用し熱エネルギを
圧力に変換して冷水を原子炉圧力容器内に注入し、冷却
材喪失事故または原子炉の緊急停止処置の必要な場合に
他の動力源を使用することなく確実に原子炉圧力容器内
に冷却水を注入することができる原子炉システムを提供
することにある。The present invention was made to solve the above problems, and uses steam discharged from a reactor pressure vessel, which does not have a structurally driven part, for example, and converts thermal energy into pressure by utilizing the latent heat of the steam. to inject cooling water into the reactor pressure vessel without using other power sources in the event of a loss of coolant accident or when emergency shutdown of the reactor is required. The goal is to provide a nuclear reactor system that can
（課題を解決するための手段）
本発明は炉心を内蔵した原子炉圧力容器と、この原子炉
圧力容器に接続された主蒸気管と、この主蒸気管に設け
られた蒸気逃し安全弁を有する蒸気逃し管と、この蒸気
逃し管および前記主蒸気管からそれぞれ分岐されたポン
プ作動系蒸気管と、このポンプ作動系蒸気管に接続され
た制御器と、前記原子炉圧力容器内へ注水する熱除去系
配管。(Means for Solving the Problems) The present invention provides a reactor pressure vessel having a built-in reactor core, a main steam pipe connected to the reactor pressure vessel, and a steam safety valve provided in the main steam pipe. a relief pipe, a pump operation system steam pipe branched from the steam relief pipe and the main steam pipe, a controller connected to the pump operation system steam pipe, and a heat removal system for injecting water into the reactor pressure vessel. System piping.
高圧注入系配管および低圧注入系配管と、これらの系統
配管にポンプを介して接続された水タンクとを具備し、
前記ポンプの少なくとも一つは前記制御器を介して作動
されかつ蒸気駆動用蒸気ポンプからなることを特徴とす
る。Equipped with high pressure injection system piping, low pressure injection system piping, and a water tank connected to these system piping via a pump,
At least one of the pumps is operated via the controller and comprises a steam pump for steam drive.
また、本発明において、前記蒸気駆動用ポンプは回転駆
動部がなく、耐圧ボディに作動蒸気供給口、給水口、ド
レンロ、高圧水吐出口の各給排出口を有し、かつ前記耐
圧ボディに高速流体をつくるノズル、蒸気と水を混合す
る混合スロートおよび高速となった水流を昇圧ディフュ
ーザを有しているものからなることを特徴とする。Further, in the present invention, the steam-driven pump does not have a rotational drive part, and has an operating steam supply port, a water supply port, a drain hole, and a high-pressure water discharge port in the pressure-resistant body, and has a high-speed water supply port in the pressure-resistant body. It is characterized by having a nozzle that creates fluid, a mixing throat that mixes steam and water, and a diffuser that increases the pressure of high-speed water flow.
（作 用）
熱除去系配管、高圧注入系配管および低圧注入系配管に
蒸気ポンプを接続する。主蒸気管または蒸気逃し配管系
から分岐してポンプ作動用配管を設け、バルブの開閉、
制御器からの信号によって前記ポンプ作動用配管から蒸
気を送り、各々の系統配管の蒸気ポンプを作動させて、
原子炉圧力容器内へ冷却水を供給し、炉心を冷却させる
。(Function) Connect the steam pump to the heat removal system piping, high pressure injection system piping, and low pressure injection system piping. Pump operating piping is installed branching off from the main steam pipe or steam relief piping system to open and close valves,
Sending steam from the pump operation piping in response to a signal from the controller to operate the steam pump of each system piping,
Cooling water is supplied into the reactor pressure vessel to cool the reactor core.
蒸気ポンプは熱エネルギを運動エネルギに変換し、昇圧
ディフューザでさらに蒸気の供給圧力より高い圧力まで
昇圧する。この昇圧によって原子炉圧力容器内に高い圧
力で冷却水を供給することができる。The steam pump converts thermal energy into kinetic energy, which is further boosted by a booster diffuser to a pressure higher than the steam supply pressure. This pressure increase allows cooling water to be supplied at high pressure into the reactor pressure vessel.
（実施例）
第１−図から第３図を参照しながら本発明に係る原子炉
システムの一実施例を説明する。(Embodiment) An embodiment of the nuclear reactor system according to the present invention will be described with reference to FIGS. 1 to 3.
第１図において符号］は原子炉圧力容器２内に配置され
た炉心を示している。原子炉圧力容器２には主蒸気管３
が接続されており、この主蒸気管３から分岐して蒸気逃
し弁１２を有する蒸気逃し配管１２ａが接続されている
。また主蒸気管３と蒸気逃し配管１２ａからそれぞれ分
岐してポンプ作動用蒸気配管１３の一端に接続される分
岐配管１３ａ、 １３ｂが設けられている。主蒸気管３
からの分岐配管１．３ｂにはバルブ１４が介在されてい
る。ポンプ作動用蒸気配管１３の他端は熱除去系配管５
ａ、高圧注入系配管８ａおよび低圧注入系配管９ａにそ
れぞれ設けられた蒸気ポンプ５ｂ、８ｂおよび９ｂに制
御器１５を介して接続されている。熱除去系配管５ａ、
高圧注入系配管８ａおよび低圧注入系配管９ｂの一端は
非常用タンク１０に、他端は原子炉圧力容器２にそれぞ
れ接続されている。熱除去系配管５ａの一端はまた再循
環系配管７の再循環ポンプ４の入口側に接続されている
。再循環系配管７と熱除去系配管５ａとの間には熱交換
器６が接続されている。高圧注入系配管８ａの途中には
他の非常用水タンク１０ａが接続され、多量の水を原子
炉圧力容器２内へ注水できるようになっている。In FIG. 1, the reference numeral "] indicates a core disposed within the reactor pressure vessel 2. In FIG. The reactor pressure vessel 2 has a main steam pipe 3
is connected to the main steam pipe 3, and a steam relief pipe 12a having a steam relief valve 12 is connected to the main steam pipe 3. Further, branch pipes 13a and 13b are provided which are branched from the main steam pipe 3 and the steam relief pipe 12a and connected to one end of the pump operating steam pipe 13, respectively. Main steam pipe 3
A valve 14 is interposed in the branch pipe 1.3b from. The other end of the steam piping 13 for pump operation is the heat removal system piping 5
a, connected via a controller 15 to steam pumps 5b, 8b, and 9b provided in the high-pressure injection system piping 8a and the low-pressure injection system piping 9a, respectively. heat removal system piping 5a,
One end of the high-pressure injection system piping 8a and the low-pressure injection system piping 9b is connected to the emergency tank 10, and the other end is connected to the reactor pressure vessel 2, respectively. One end of the heat removal system piping 5a is also connected to the inlet side of the recirculation pump 4 of the recirculation system piping 7. A heat exchanger 6 is connected between the recirculation system piping 7 and the heat removal system piping 5a. Another emergency water tank 10a is connected in the middle of the high-pressure injection system piping 8a, so that a large amount of water can be injected into the reactor pressure vessel 2.
−２で、各系統配管５ａ、８ａおよび９ｂに組込まれる
蒸気ポンプ５ｂ、 ８ｂおよび９ｂは第２図に示したと
同様の構造を有する蒸気ポンプ１１である。-2, the steam pumps 5b, 8b and 9b incorporated in each system piping 5a, 8a and 9b are steam pumps 11 having the same structure as shown in FIG.
すなわち、第２図に示した蒸気ポンプ１１は供給する水
の圧力に十分耐え得る耐圧ボディ２０に作動蒸気供給口
１６（１７）と、これに対向して給水口１７（１６）が
設けられている。またドレン口１８がボディ２０の下側
面に、高圧水吐出口１９がボディ２０の右端に接続され
ている。さらにボディ２０内には作動蒸気供給口１６（
１７）から流入した蒸気を絞って噴出させるノズル２１
と、このノズル２１からの蒸気と給水口１７（１６）か
らの水とを混合する混合スロート２２と、高圧水吐出口
１９側に設けられた昇圧ディフューザ２３とを有してい
る。なお、蒸気供給口１６の位置と給水口１７の位置を
逆にすることもできる。That is, the steam pump 11 shown in FIG. 2 is provided with an operating steam supply port 16 (17) and a water supply port 17 (16) opposite to this in a pressure-resistant body 20 that can sufficiently withstand the pressure of supplied water. There is. Further, the drain port 18 is connected to the lower side of the body 20, and the high-pressure water discharge port 19 is connected to the right end of the body 20. Further, inside the body 20 is an operating steam supply port 16 (
17) Nozzle 21 that squeezes and spouts the steam flowing in from
, a mixing throat 22 that mixes the steam from the nozzle 21 and water from the water supply port 17 (16), and a pressure boosting diffuser 23 provided on the high-pressure water discharge port 19 side. Note that the positions of the steam supply port 16 and the water supply port 17 can also be reversed.
しかして、蒸気ポンプ１１は混合スロート２２で高温の
蒸気と低温の水を混合することで蒸気の熱エネルギを水
の運動エネルギと圧力に変換し、昇圧ディフューザ２３
でさらに蒸気の供給圧力より高い圧力まで昇圧する。こ
れによって、原子炉圧内容器２内の熱を利用し原子炉圧
力容器２内の圧力より高い圧力で冷却水を原子炉圧力容
器２内に供給することができる。Thus, the steam pump 11 mixes high-temperature steam and low-temperature water at the mixing throat 22 to convert the thermal energy of the steam into kinetic energy and pressure of water, and the pressure boosting diffuser 23
The pressure is further increased to a pressure higher than the steam supply pressure. Thereby, cooling water can be supplied into the reactor pressure vessel 2 at a pressure higher than the pressure inside the reactor pressure vessel 2 by utilizing the heat within the reactor pressure vessel 2 .
第３図は蒸気ポンプ１１の作用を説明するためのもので
、第２図と同一部分には同一符号で示し重複する部分の
説明を省略する。FIG. 3 is for explaining the operation of the steam pump 11, and the same parts as in FIG. 2 are denoted by the same reference numerals, and the explanation of the overlapping parts will be omitted.
第３図中、■を蒸気入口、■を水供給口、■を吐出口と
し、また、式中、Ｑを水９ｇを蒸気２ｍを流量、ｈをエ
ンタルピー、Ｐを圧力、ρを密度とする。すると。In Figure 3, ■ is the steam inlet, ■ is the water supply port, and ■ is the discharge port, and in the formula, Q is the flow rate of 9 g of water and 2 m of steam, h is the enthalpy, P is the pressure, and ρ is the density. . Then.
このように蒸気ポンプ１１は蒸気の潜熱を利用して熱エ
ネルギを圧力に変換して冷却水を原子炉圧力容器内に注
入できる。したがって、とくに大きな容量を必要としな
いポンプにこの蒸気ポンプを適用することで駆動部のな
い信頼性の高いポンプを提供できる。In this way, the steam pump 11 can convert thermal energy into pressure using the latent heat of steam and inject cooling water into the reactor pressure vessel. Therefore, by applying this steam pump to a pump that does not require a particularly large capacity, a highly reliable pump without a driving part can be provided.
上記実施例によれば原子炉の常用、非常用の冷却水供給
系に蒸気ポンプを組込むことによって、冷却材喪失事故
または原子炉の緊急停止処置の必要な場合に他の動力源
を使用しないで、確実に原子炉圧力容器内に注入でき、
また系統配管のポンプその他の機器のメンテナンスを容
易にすることができる。According to the above embodiment, by incorporating a steam pump into the reactor's regular and emergency cooling water supply systems, other power sources are not used in the event of a loss of coolant accident or when an emergency shutdown of the reactor is required. , can be reliably injected into the reactor pressure vessel,
Furthermore, maintenance of pumps and other equipment in the system piping can be facilitated.
なお、本発明の実施態様を要約すればつぎのとうりであ
る。■、原子炉圧力容器の蒸気を配管で蒸気ポンプに導
き、低圧の低温水を加圧して原子炉圧力容器の圧力より
高い圧力で原子炉圧力容器内へ注入すること。■蒸気供
給系に開閉弁（バルブ）、流量調整器を設けて蒸気ポン
プの作動を制御すること。■非常時には蒸気逃し管など
制御されない系統配管から蒸気を供給する非常用給水系
をもつこと。＠）蒸気ポンプは回転機構などの駆動部が
なく、作動蒸気供給口、給水口、ドレンロ、高圧水吐出
口の各給排出口を有し、供給する水の圧力に十分耐える
耐圧ボディと、高速流体をつくるノズルと、蒸気と水と
を混合する混合スロー１〜、さらに高速となった水流を
昇圧する昇圧ディフューザとからなることである。The embodiments of the present invention can be summarized as follows. (2) Leading steam from the reactor pressure vessel to a steam pump via piping, pressurizing low-pressure low-temperature water, and injecting it into the reactor pressure vessel at a pressure higher than that of the reactor pressure vessel. ■The steam supply system must be equipped with an on-off valve and a flow rate regulator to control the operation of the steam pump. ■Have an emergency water supply system that supplies steam from uncontrolled system piping such as steam relief pipes in the event of an emergency. @) A steam pump does not have a rotating mechanism or other driving part, but has an operating steam supply port, a water supply port, a drain hole, and a high-pressure water discharge port.It has a pressure-resistant body that can withstand the pressure of the water it supplies, and has a high-speed pump. It consists of a nozzle that creates fluid, a mixing throw 1 to mix steam and water, and a pressure boosting diffuser that boosts the pressure of the high-speed water flow.
本発明によれば、炉心から発生する余剰の蒸気を用いて
蒸気ポンプを作動させることで水タンクに溜められた常
圧の水を加圧して、供給された蒸気圧よりも高い圧力と
して原子炉圧力容器内に供給することができ、原子炉停
止時の原子炉圧力容器内の熱を除去することができる。According to the present invention, by operating a steam pump using excess steam generated from the reactor core, normal pressure water stored in a water tank is pressurized, and the pressure is increased to a higher pressure than the supplied steam pressure. It can be supplied into the pressure vessel, and the heat inside the reactor pressure vessel when the reactor is shut down can be removed.
すなわち、原子炉圧力容器内への冷却材注入を外部から
の動力の供給をせずに、また稼動部のないポンプを用い
ることで原子炉圧力容器内からの発熱による蒸気の供給
と供給する水がある限り、原子炉圧力容器内への冷却水
の供給が可能となり、安定した信頼性の高い原子炉シス
テムを得る。In other words, by injecting coolant into the reactor pressure vessel without supplying power from outside, and by using a pump with no moving parts, it is possible to supply steam and water by generating heat from inside the reactor pressure vessel. As long as there is water available, cooling water can be supplied into the reactor pressure vessel, resulting in a stable and highly reliable nuclear reactor system.
第１図は本発明に係る原子炉システムの一実施例を示す
系統図、第２図は第１図の要部における蒸気ポンプを示
す縦断面図、第３図は第２図における蒸気ポンプの原理
を示す縦断面図、第４図は従来の原子炉システムの冷却
水注入系を示す系統図である。
１・・・炉心 ２・・・原子炉圧力容器
３・・・主蒸気管 ４ 再循環ポンプ５ ・
熱除去用ポンプ ５ａ・熱除去系配管６・・・熱交
換器 ７・・再循環系配管８・・・高圧注入
用ポンプ ８ａ・高圧注入系配管９・・・低圧ポンプ
用ポンプ ９ａ・・低圧注入系配管１０・・非常用水タ
ンク １１・・蒸気ポンプ１２・・・逃し弁
１３・・ポンプ作動用蒸気配管１４・バルブ
１５・・・制御器１６・・・作動蒸気供給
口 １７・給水口１８・・ドレン口 １
９・・高圧水吐出口２０・・耐圧ボディ ２１
・・ノズル２２・・混合スロート ２３・・昇圧
ディフューザ（８７３３）代理人弁理士 猪 股 祥
晃（ばか１名）第
図FIG. 1 is a system diagram showing an embodiment of the nuclear reactor system according to the present invention, FIG. 2 is a vertical cross-sectional view showing the steam pump in the main part of FIG. 1, and FIG. 3 is a schematic diagram of the steam pump in FIG. FIG. 4 is a longitudinal sectional view showing the principle, and is a system diagram showing a cooling water injection system of a conventional nuclear reactor system. 1... Reactor core 2... Reactor pressure vessel 3... Main steam pipe 4 Recirculation pump 5 ・
Heat removal pump 5a, heat removal system piping 6...heat exchanger 7, recirculation system piping 8...high pressure injection pump 8a, high pressure injection system piping 9...low pressure pump pump 9a...low pressure Injection system piping 10... Emergency water tank 11... Steam pump 12... Relief valve
13. Steam piping for pump operation 14. Valve
15...Controller 16...Working steam supply port 17.Water supply port 18...Drain port 1
9...High pressure water outlet 20...Pressure resistant body 21
... Nozzle 22 ... Mixing throat 23 ... Boosting diffuser (8733) Patent attorney Sho Inomata
Akira (one idiot) diagram
Claims (2)
力容器に接続された主蒸気管と、この主蒸気管に設けら
れた蒸気逃し安全弁を有する蒸気逃し管と、この蒸気逃
し管および前記主蒸気管からそれぞれ分岐されたポンプ
作動系蒸気管と、このポンプ作動系蒸気管に接続された
制御器と、前記原子炉圧力容器内へ注水する熱除去系配
管、高圧注入系配管および低圧注入系配管と、これらの
系統配管にポンプを介して接続された水タンクとを具備
し、前記ポンプの少なくとも一つは前記制御器を介して
作動されかつ蒸気駆動用蒸気ポンプからなることを特徴
とする原子炉システム。(1) A reactor pressure vessel containing a reactor core, a main steam pipe connected to this reactor pressure vessel, a steam relief pipe provided in this main steam pipe having a steam relief safety valve, and this steam relief pipe and A pump operation system steam pipe branched from the main steam pipe, a controller connected to the pump operation system steam pipe, a heat removal system piping for injecting water into the reactor pressure vessel, a high pressure injection system piping, and a low pressure It comprises injection system piping and a water tank connected to these system piping via a pump, and at least one of the pumps is operated via the controller and comprises a steam pump for steam drive. nuclear reactor system.
ボディに作動蒸気供給口、給水口、ドレン口、高圧水吐
出口の各給排出口を有し、かつ前記耐圧ボディに高速流
体をつくるノズル、蒸気と水を混合する混合スロートお
よび高速となった水流を昇圧ディフューザを有している
ものからなることを特徴とする請求項１記載の原子炉シ
ステム。(2) The steam-driven pump does not have a rotational drive part, has a pressure-resistant body with an operating steam supply port, a water supply port, a drain port, and a high-pressure water discharge port, and has a high-speed fluid in the pressure-resistant body. 2. The nuclear reactor system according to claim 1, further comprising a nozzle for mixing steam and water, a mixing throat for mixing steam and water, and a diffuser for boosting the pressure of the high-speed water flow.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1073925A JP2744056B2 (en) | 1989-03-28 | 1989-03-28 | Cooling water injection system for boiling water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP1073925A JP2744056B2 (en) | 1989-03-28 | 1989-03-28 | Cooling water injection system for boiling water reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPH02253194A true JPH02253194A (en) | 1990-10-11 |
| JP2744056B2 JP2744056B2 (en) | 1998-04-28 |
Family
ID=13532208
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP1073925A Expired - Fee Related JP2744056B2 (en) | 1989-03-28 | 1989-03-28 | Cooling water injection system for boiling water reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JP2744056B2 (en) |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS6355493A (en) * | 1986-08-27 | 1988-03-09 | 株式会社日立製作所 | Heavy water moderation pressure tube type reactor |
-
1989
- 1989-03-28 JP JP1073925A patent/JP2744056B2/en not_active Expired - Fee Related
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS6355493A (en) * | 1986-08-27 | 1988-03-09 | 株式会社日立製作所 | Heavy water moderation pressure tube type reactor |
Also Published As
| Publication number | Publication date |
|---|---|
| JP2744056B2 (en) | 1998-04-28 |
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