JPH06230177A - Boiling water reactor - Google Patents
Boiling water reactorInfo
- Publication number
- JPH06230177A JPH06230177A JP5015790A JP1579093A JPH06230177A JP H06230177 A JPH06230177 A JP H06230177A JP 5015790 A JP5015790 A JP 5015790A JP 1579093 A JP1579093 A JP 1579093A JP H06230177 A JPH06230177 A JP H06230177A
- Authority
- JP
- Japan
- Prior art keywords
- reactor
- pressure
- cooling system
- vessel
- nuclear 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.)
- Pending
Links
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
-
- 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)
Abstract
Description
【０００１】[0001]
【産業上の利用分野】本発明は、沸騰水型原子炉に係
り、特に原子力プラントの全交流電源喪失事故に際して
原子炉圧力容器を安全に保つ沸騰水型原子炉に関する。BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a boiling water reactor, and more particularly to a boiling water reactor that keeps a reactor pressure vessel safe in the event of a loss of all AC power supplies in a nuclear power plant.
【０００２】[0002]
【従来の技術】一般に、沸騰水型原子炉においては、何
らかの原因により外部電源が喪失し、かつ、非常用ディ
ーゼル発電機による給電にも失敗するような全交流電源
喪失事故が発生した場合においても、原子炉施設の安全
確保に必要な機器へ直流電源設備から電力を供給し、原
子炉の緊急停止を行うと共に炉心を冷却して、原子力プ
ラントを安全な状態に保つことができるように構成され
ている。2. Description of the Related Art Generally, in a boiling water nuclear reactor, even if an external power source is lost due to some cause and an AC power loss accident occurs in which power supply from an emergency diesel generator also fails. , Which is configured to supply power from the DC power supply equipment to equipment necessary for ensuring the safety of the nuclear reactor facility, perform an emergency shutdown of the reactor, cool the core, and keep the nuclear power plant in a safe state. ing.
【０００３】図２の要部系統構成図は従来の沸騰水型原
子炉を示したもので、原子炉格納容器１内に設置されて
いる原子炉圧力容器２には核燃料が装荷された炉心３が
収容されていて、核エネルギーを得た冷却水は蒸気とな
り主蒸気管４を介して図示しない発電用タービンに導か
れる。発電用タービンを駆動した蒸気は図示しない復水
器によって復水とされ、給水系によって再び原子炉圧力
容器２に戻される。FIG. 2 is a system diagram of a main part of a conventional boiling water reactor. A reactor pressure vessel 2 installed in a reactor containment vessel 1 has a reactor core 3 loaded with nuclear fuel. The cooling water which has received the nuclear energy becomes steam and is guided to a power generation turbine (not shown) through the main steam pipe 4. The steam that has driven the power generation turbine is condensed by a condenser (not shown) and returned to the reactor pressure vessel 2 again by the water supply system.
【０００４】この沸騰水型原子炉において、全交流電源
喪失事故が発生した場合には、別途設置された図示しな
い直流電源設備によって原子炉の安全運転に必要な機器
に対して給電が開始されることにより、原子炉は緊急停
止され、復水、給水系は停止して復水器から隔離状態に
おかれる。In this boiling water reactor, when an accident of loss of all AC power supplies occurs, power supply to equipment required for safe operation of the reactor is started by a separately installed DC power supply facility (not shown). As a result, the reactor is shut down urgently, and the condensate and water supply systems are stopped and the reactor is isolated from the condenser.
【０００５】この際に炉心燃料の崩壊熱によって発生し
た蒸気は原子炉圧力を上昇させるため、原子炉圧力容器
２内の蒸気を主蒸気逃し安全弁５を介してサプレッショ
ンプール６の水中へ放出する。これにより原子炉圧力は
主蒸気逃し安全弁５の設定圧力（約75kg／cm2 ｇ）程度
に一定に保たれる。At this time, the steam generated by the decay heat of the core fuel raises the reactor pressure, so that the steam in the reactor pressure vessel 2 is released into the water in the suppression pool 6 via the main steam relief valve 5. As a result, the reactor pressure is set to the main steam relief safety valve 5 (about 75 kg / cm 2 g) to be kept constant.
【０００６】また復水、給水系が停止したことにより原
子炉水位は低下するが、原子炉３で発生する蒸気の一部
を用いた蒸気駆動タービン７で駆動されるタービン駆動
ポンプ８により、復水貯蔵槽９あるいはサプレッション
プール６内の水を原子炉圧力容器２に注水する原子炉隔
離時冷却系が自動起動して、原子炉の水位回復を図り、
十分な炉心冷却が行われる。なお、原子炉隔離時冷却系
においては、原子炉圧力が約80〜10kg／cm2 ｇの範囲で
原子炉への給水が可能なように設計されている。Although the reactor water level is lowered due to the stop of the condensate and water supply system, the turbine drive pump 8 driven by the steam drive turbine 7 using a part of the steam generated in the reactor 3 The cooling system at the time of reactor isolation for automatically injecting the water in the water storage tank 9 or the suppression pool 6 into the reactor pressure vessel 2 is automatically started, and the water level of the reactor is restored.
Sufficient core cooling is performed. In the reactor isolation cooling system, the reactor pressure is approximately 80 to 10 kg / cm 2 It is designed to supply water to the reactor in the range of g.
【０００７】以上のように全交流電源喪失事故において
は、交流電源に依存せず直流電源設備からの給電によっ
て機能する原子炉隔離時冷却系と主蒸気逃し安全弁５の
作動によって、原子炉を安定な状態に維持することが可
能となっている。As described above, in the event of a loss of all AC power, the reactor isolation cooling system and the main steam relief safety valve 5 that function by the power supply from the DC power supply without depending on the AC power are operated to stabilize the reactor. It is possible to maintain a good state.
【０００８】また直流電源設備は、全交流電源喪失事故
発生後に必要な機器に対して数時間の給電が可能な運転
時間容量が確保されており、この間に外部電源の復旧が
行われ、あるいは非常用ディーゼル発電機からの給電に
より、原子力プラントの安全は確保されて電源喪失によ
る事故は収束される。In addition, the DC power supply equipment has an operating time capacity capable of supplying power for several hours to the necessary equipment after the occurrence of the loss of all AC power supplies. During this time, the external power supply is restored or emergency By supplying power from the diesel power generator, the safety of the nuclear power plant is secured and the accident caused by the loss of power source is resolved.
【０００９】[0009]
【発明が解決しようとする課題】沸騰水型原子炉におい
て全交流電源喪失事故が発生した際に、若しも直流電源
設備の運転時間容量内に外部電源が復旧せず、かつ非常
用ディーゼル発電機からの給電も復旧しない場合には、
いずれ直流電源は枯渇する。この場合には原子炉隔離時
冷却系による炉心３への注水が停止し、原子炉の水位が
低下し始める。しかも、万一、その後においても電源復
旧が行われない場合には、炉心３は蒸気雰囲気中に徐々
に露出し始め、事故の崩壊熱および水−金属反応による
発熱によって過熱され、炉心３が損傷する事態へと進展
して、さらには原子炉圧力容器２が損傷した炉心２によ
って破損することも想定される。When a loss of all AC power supply occurs in a boiling water reactor, the external power supply cannot be restored within the operating time capacity of the DC power supply equipment, and emergency diesel power generation is not possible. If the power supply from the machine is not restored,
DC power will eventually be exhausted. In this case, the injection of water into the core 3 by the reactor isolation cooling system is stopped, and the water level of the reactor begins to drop. Moreover, if the power supply is not restored even after that, the core 3 gradually begins to be exposed in the steam atmosphere and is overheated due to the decay heat of the accident and the heat generated by the water-metal reaction, and the core 3 is damaged. It is assumed that the reactor pressure vessel 2 is further damaged by the damaged core 2.
【００１０】この場合に原子炉隔離時冷却系によって炉
心冷却が行われている期間では、原子炉圧力は５の設定
圧力近傍の約75kg／cm2 ｇに維持されるため、原子炉圧
力容器２は長時間高圧状態となり破損することが予想さ
れ、続いて原子炉格納容器１の早期破損に結びつく可能
性がある。このような事故が発生することは確率的には
極めて低いものであるが、万一発生した場合には公衆へ
の多大な被害を及ぼすことが考えられる。In this case, during the period in which core cooling is performed by the reactor isolation cooling system, the reactor pressure is about 75 kg / cm 2 near the set pressure of 5. Since the reactor pressure vessel 2 is maintained at g, it is expected that the reactor pressure vessel 2 will be in a high pressure state for a long time and will be damaged, which may lead to early failure of the reactor containment vessel 1. Occurrence of such an accident is extremely low in probability, but if it should occur, it may cause a great deal of damage to the public.
【００１１】また原子炉圧力容器２の高圧破損を防止す
るために、従来の自動減圧機能を有する主蒸気逃し安全
弁５を手動操作により原子炉を不用意に減圧した場合に
は、原子炉圧力が原子炉隔離時冷却系の運転最低圧力以
下に減圧されるため、原子炉隔離時冷却系による炉心冷
却ができず、炉心損傷が早まる恐れもある。Further, in order to prevent the high pressure damage of the reactor pressure vessel 2, if the main steam relief safety valve 5 having a conventional automatic depressurization function is inadvertently depressurized, the reactor pressure is Since the pressure is reduced below the minimum operating pressure of the reactor isolation cooling system, core cooling cannot be performed by the reactor isolation cooling system, and core damage may be accelerated.
【００１２】本発明の目的とするところは、全交流電源
喪失事故時において、主蒸気逃し安全弁により原子炉圧
力を原子炉隔離時冷却系の運転を損わない程度まで自動
的に減圧して一定状態を維持し、万一、炉心損傷に至り
原子炉圧力容器が破損した場合でも、原子炉圧力容器の
高圧破損がなく安全性の高い沸騰水型原子炉を提供する
ことにある。The object of the present invention is to automatically reduce the reactor pressure by the main steam relief safety valve to the extent that it does not impair the operation of the reactor isolation cooling system in the event of a loss of all AC power supplies. (EN) Provided is a highly safe boiling water reactor in which the state is maintained and even if the reactor pressure vessel is damaged and the reactor pressure vessel is damaged, the reactor pressure vessel is not damaged by high pressure.
【００１３】[0013]
【課題を解決するための手段】原子炉格納容器内に炉心
を収容した原子炉圧力容器を設置して原子炉で発生する
蒸気を駆動源としたポンプにより冷却水を原子炉へ供給
する原子炉隔離時冷却系を備えてなる沸騰水型原子炉に
おいて、前記原子炉圧力容器に原子炉圧力の過大な上昇
を防止する主蒸気逃し安全弁を設置すると共に、この主
蒸気逃し安全弁の設定圧力を前記原子炉隔離時冷却系の
作動に際して原子炉隔離時冷却系の運転最低圧力よりも
高い設定圧力に変更する制御装置を設けたことを特徴と
する。Reactor for supplying cooling water to a reactor by a reactor pressure vessel containing a reactor core in a reactor containment vessel and pumping steam generated in the reactor as a drive source. In a boiling water reactor equipped with an isolation cooling system, a main steam relief safety valve that prevents an excessive rise in reactor pressure is installed in the reactor pressure vessel, and the set pressure of this main steam relief safety valve is It is characterized in that a control device is provided for changing the set pressure higher than the minimum operating pressure of the reactor isolation cooling system when the reactor isolation cooling system is operated.
【００１４】[0014]
【作用】本発明によれば、全交流電源喪失事故等で原子
炉隔離時冷却系が作動した場合に、制御装置により主蒸
気逃し安全弁の設定圧力が従来の主蒸気逃し安全弁の設
定圧力より低く、原子炉隔離時冷却系の運転最低圧力よ
りも高めの圧力に変更される。According to the present invention, when the reactor isolation cooling system is activated due to an accident such as loss of all AC power, the controller sets the main steam relief safety valve to have a lower set pressure than the conventional main steam relief safety valve. , The pressure will be changed to a pressure higher than the minimum operating pressure of the reactor isolation cooling system.
【００１５】これにより、全交流電源喪失事故時におい
て原子炉圧力は、蒸気をサプレッションプールの水中に
排出することにより、その設定圧力まで自動的に低下さ
れて一定に維持される。さらに、直流電源の枯渇および
非常用ディーゼル発電機の始動失敗が重なるような場合
で炉心損傷に至る際にも、原子炉圧力が低下しているた
め原子炉圧力容器の高圧破損は防止され、これに付随し
て原子炉格納容器の早期破損が防止される。As a result, in the event of a loss of all AC power supplies, the reactor pressure is automatically reduced to the set pressure by steam being discharged into the water of the suppression pool, and is kept constant. Furthermore, when the direct current power supply is exhausted and the emergency diesel generator fails to start, and the reactor core is damaged, the high pressure damage of the reactor pressure vessel is prevented because the reactor pressure is reduced. Due to this, early damage of the reactor containment vessel is prevented.
【００１６】[0016]
【実施例】本発明の一実施例を図面を参照して説明す
る。なお、上記した従来技術と同じ構成部分については
同一符号を付して詳細な説明を省略する。DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to the drawings. It should be noted that the same components as those in the above-described conventional technique are designated by the same reference numerals and detailed description thereof will be omitted.
【００１７】図１の要部系統構成図に示すように、原子
炉格納容器１内に設置されている原子炉圧力容器２には
炉心３が収容されていて、核エネルギーを得た冷却水は
蒸気となり主蒸気管４を介して図示しない発電用タービ
ンに導かれる。発電用タービンを駆動した蒸気は図示し
ない復水器によって復水とされ、給水系によって再び原
子炉圧力容器２に戻される。As shown in the main system configuration diagram of FIG. 1, the reactor core 3 is housed in the reactor pressure vessel 2 installed in the reactor containment vessel 1, and the cooling water that has obtained nuclear energy is It becomes steam and is guided to a power generation turbine (not shown) through the main steam pipe 4. The steam that has driven the power generation turbine is condensed by a condenser (not shown) and returned to the reactor pressure vessel 2 again by the water supply system.
【００１８】また原子炉圧力容器２には内部の圧力が異
常に上昇した場合に、原子炉圧力容器２の蒸気をサプレ
ッションプール６の水中へ放出して原子炉圧力容器２の
安全を確保する主蒸気逃し安全弁10が設置されていて、
この主蒸気逃し安全弁10が作動する設定圧力は、制御装
置11により通常運転時は約75kg／cm2 ｇに、また原子炉
隔離時冷却系の動作時には、この原子炉隔離時冷却系の
運転最低圧力（約10kg／cm2 ｇ）よりも幾分高めの圧力
（例えば約15kg／cm2 ｇ）に設定値が変更されるように
構成されている。次に上記構成による作用について説明
する。Further, when the internal pressure of the reactor pressure vessel 2 rises abnormally, the steam of the reactor pressure vessel 2 is discharged into the water of the suppression pool 6 to ensure the safety of the reactor pressure vessel 2. A steam relief valve 10 is installed,
The set pressure at which this main steam relief safety valve 10 operates is about 75 kg / cm 2 during normal operation by the control device 11. g, and when operating the reactor isolation cooling system, the minimum operating pressure of this reactor isolation cooling system (about 10 kg / cm 2 g) somewhat higher pressure (eg about 15 kg / cm 2 The setting value is changed in g). Next, the operation of the above configuration will be described.
【００１９】原子炉は通常運転時においては上記した従
来と同様に運転されているが、全交流電源喪失事故が発
生した場合には、別途設置された図示しない直流電源設
備によって必要な機器に給電が開始されると共に原子炉
は緊急停止され、復水、給水系は停止して復水器から隔
離状態におかれる。In normal operation, the nuclear reactor is operated in the same manner as the above-mentioned conventional one. However, in the event of a total AC power loss accident, a separately installed DC power supply facility supplies power to necessary equipment. With the start of the reactor, the reactor is shut down urgently, and the condensate and water supply systems are stopped and the reactor is isolated from the condenser.
【００２０】また復水、給水系が停止したことにより原
子炉水位は低下するが、原子炉隔離時冷却系である原子
炉３で発生する蒸気の一部を用いた蒸気駆動タービン７
によって駆動するタービン駆動ポンプ８により、復水貯
蔵槽９あるいはサプレッションプール６の水を原子炉圧
力容器２に注水する原子炉隔離時冷却系が自動起動して
原子炉の水位回復を図り、十分な炉心冷却が行われる。Further, although the reactor water level is lowered due to the stop of the condensate and water supply systems, the steam driven turbine 7 using a part of the steam generated in the reactor 3 which is the cooling system during reactor isolation.
By the turbine drive pump 8 driven by the reactor, the reactor isolation cooling system for injecting the water in the condensate storage tank 9 or the suppression pool 6 into the reactor pressure vessel 2 is automatically started to restore the water level in the reactor sufficiently. Core cooling is performed.
【００２１】この場合に原子炉圧力容器２内で炉心燃料
の崩壊熱によって発生した蒸気は、主蒸気逃し安全弁10
を介してサプレッションプール６の水中へ放出される
が、原子炉隔離時冷却系の起動と共に主蒸気逃し安全弁
10が作動する設定圧力が、制御装置11により原子炉隔離
時冷却系の運転最低圧力（約10kg／cm2 ｇ）よりも幾分
高めの圧力（例えば約15kg／cm2 ｇ）に設定変更され
る。In this case, the steam generated by the decay heat of the core fuel in the reactor pressure vessel 2 is the main steam relief safety valve 10
Is released into the water of the suppression pool 6 via the main steam release safety valve when the reactor isolation cooling system is started.
The set pressure at which 10 operates is determined by the controller 11 as the minimum operating pressure of the cooling system during reactor isolation (about 10 kg / cm 2 g) somewhat higher pressure (eg about 15 kg / cm 2 The setting is changed to g).
【００２２】これにより原子炉は所定の圧力まで減圧さ
れ、全交流電源喪失事故時専用に設定された一定圧力に
維持され、原子炉は従来と同様に原子炉隔離時冷却系に
よって冷却される。As a result, the reactor is depressurized to a predetermined pressure and maintained at a constant pressure set exclusively for the loss of all AC power, and the reactor is cooled by the reactor isolation cooling system as in the conventional case.
【００２３】なお、万一、原子炉隔離時冷却系等による
炉心冷却を維持する機器へ給電する直流電源設備の運転
時間容量以後も外部電源が復旧せず、あるいは非常用デ
ィーゼル発電機からの電力供給が行われない場合を想定
すると、その結果、炉心３が損傷して原子炉圧力容器２
が破損に至るような事故が発生した場合においても、原
子炉圧力容器２の圧力が低く維持されているため、原子
炉圧力容器２の高圧破損は発生しないことから、原子炉
格納容器１の健全性早期喪失を防止することができる。In the unlikely event that the external power supply is not restored even after the operation time capacity of the DC power supply facility for supplying power to the equipment for maintaining the core cooling by the reactor isolation cooling system or the like, the power from the emergency diesel generator is used. Assuming that the supply is not performed, as a result, the core 3 is damaged and the reactor pressure vessel 2 is damaged.
Even in the event of an accident that results in damage to the reactor, the high pressure of the reactor pressure vessel 2 does not occur because the pressure of the reactor pressure vessel 2 is kept low. Premature loss can be prevented.
【００２４】[0024]
【発明の効果】以上本発明によれば、長期間の全交流電
源喪失事故が発生し、かつ炉心が損傷するような事故に
至った場合においても、原子炉圧力容器の高圧破損が防
止されることから、原子炉格納容器健全性の早期喪失も
防止され、沸騰水型原子力プラントの安全性が向上する
効果がある。As described above, according to the present invention, even if an accident such as loss of all AC power for a long time occurs and the core is damaged, high pressure damage of the reactor pressure vessel is prevented. Therefore, the early loss of soundness of the reactor containment vessel is prevented, and the safety of the boiling water nuclear power plant is improved.
【図１】本発明に係る一実施例を示す沸騰水型原子炉の
要部系統構成図。FIG. 1 is a system configuration diagram of a main part of a boiling water reactor showing an embodiment according to the present invention.
【図２】従来の沸騰水型原子炉の要部系統構成図。FIG. 2 is a system configuration diagram of a main part of a conventional boiling water reactor.
１…原子炉格納容器、２…原子炉圧力容器、３…炉心、
４…主蒸気管、５，10…主蒸気逃し安全弁、６…サプレ
ッションプール、７…蒸気駆動タービン、８…タービン
駆動ポンプ、９…復水貯蔵槽、11…制御装置。1 ... Reactor containment vessel, 2 ... Reactor pressure vessel, 3 ... Reactor core,
4 ... Main steam pipe, 5, 10 ... Main steam relief safety valve, 6 ... Suppression pool, 7 ... Steam driven turbine, 8 ... Turbine driven pump, 9 ... Condensate storage tank, 11 ... Control device.
Claims (1)
炉圧力容器を設置すると共に原子炉で発生する蒸気を駆
動源としたポンプにより冷却水を原子炉へ供給する原子
炉隔離時冷却系を備えてなる沸騰水型原子炉において、
前記原子炉圧力容器に原子炉圧力の過大な上昇を防止す
る主蒸気逃し安全弁を設置すると共に、この主蒸気逃し
安全弁の設定圧力を前記原子炉隔離時冷却系の作動に際
して原子炉隔離時冷却系の運転最低圧力よりも高い設定
圧力に変更する制御装置を設けたことを特徴とする沸騰
水型原子炉。1. A reactor isolation cooling system in which a reactor pressure vessel containing a reactor core is installed in a reactor containment vessel, and cooling water is supplied to the reactor by a pump driven by steam generated in the reactor. In a boiling water reactor comprising
A main steam relief safety valve is installed in the reactor pressure vessel to prevent an excessive rise in the reactor pressure, and the set pressure of the main steam relief safety valve is set to the reactor isolation cooling system when the reactor isolation cooling system is operated. A boiling water reactor characterized by being provided with a control device for changing the set pressure to a value higher than the minimum operating pressure of.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5015790A JPH06230177A (en) | 1993-02-03 | 1993-02-03 | Boiling water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP5015790A JPH06230177A (en) | 1993-02-03 | 1993-02-03 | Boiling water reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPH06230177A true JPH06230177A (en) | 1994-08-19 |
Family
ID=11898638
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP5015790A Pending JPH06230177A (en) | 1993-02-03 | 1993-02-03 | Boiling water reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPH06230177A (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2014055948A (en) * | 2012-09-11 | 2014-03-27 | Ge-Hitachi Nuclear Energy Americas Llc | Method and system for external alternate suppression pool cooling for boiling water nuclear reactor |
| JP2016105058A (en) * | 2014-12-01 | 2016-06-09 | 日立Ｇｅニュークリア・エナジー株式会社 | Reactor core isolation cooling system |
| CN113744902A (en) * | 2021-07-22 | 2021-12-03 | 中国核动力研究设计院 | Natural circulation cooling method for preventing upper seal head of pressure vessel from generating steam in nuclear power plant |
-
1993
- 1993-02-03 JP JP5015790A patent/JPH06230177A/en active Pending
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JP2014055948A (en) * | 2012-09-11 | 2014-03-27 | Ge-Hitachi Nuclear Energy Americas Llc | Method and system for external alternate suppression pool cooling for boiling water nuclear reactor |
| US10395784B2 (en) | 2012-09-11 | 2019-08-27 | Ge-Hitachi Nuclear Energy Americas Llc | Method and system for external alternate suppression pool cooling for a BWR |
| JP2016105058A (en) * | 2014-12-01 | 2016-06-09 | 日立Ｇｅニュークリア・エナジー株式会社 | Reactor core isolation cooling system |
| CN113744902A (en) * | 2021-07-22 | 2021-12-03 | 中国核动力研究设计院 | Natural circulation cooling method for preventing upper seal head of pressure vessel from generating steam in nuclear power plant |
| CN113744902B (en) * | 2021-07-22 | 2023-11-24 | 中国核动力研究设计院 | Natural circulation cooling method for preventing steam generation of upper end enclosure of pressure vessel in nuclear power plant |
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