JP4607681B2 - Reactor containment equipment and pressure control method thereof - Google Patents

Reactor containment equipment and pressure control method thereof Download PDF

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JP4607681B2
JP4607681B2 JP2005180241A JP2005180241A JP4607681B2 JP 4607681 B2 JP4607681 B2 JP 4607681B2 JP 2005180241 A JP2005180241 A JP 2005180241A JP 2005180241 A JP2005180241 A JP 2005180241A JP 4607681 B2 JP4607681 B2 JP 4607681B2
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寿敏 白濱
正義 松浦
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Hitachi GE Nuclear Energy Ltd
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Description

本発明は原子力発電プラントにおける原子炉格納設備及びその圧力制御方法に係り、特に、原子炉圧力容器が収容された上部空間とサプレッションプールを形成した下部空間に真空破壊弁を備えた原子炉格納設備及びその圧力制御方法に関する。   The present invention relates to a reactor containment facility and a pressure control method thereof in a nuclear power plant, and in particular, a reactor containment facility provided with a vacuum break valve in an upper space in which a reactor pressure vessel is accommodated and a lower space in which a suppression pool is formed. And a pressure control method thereof.

従来の原子力発電プラントにおける原子炉格納設備は、冷却材喪失事故時に生じる原子炉圧力容器が収容された上部空間（ドライウェル）の圧力上昇や水頭圧を利用して原子炉格納容器の冷却及び炉心へ冷却水を注水する静的非常炉心注水系が備えてある。そして、単純沸騰水型発電設備の静的非常炉心注水系として、静的格納容器冷却系と重力落下式炉心冷却系がある。   The conventional containment facilities in nuclear power plants use a rise in pressure in the upper space (dry well) containing the reactor pressure vessel (dry well) generated during a loss of coolant or water head pressure to cool the reactor containment vessel and the core. A static emergency core water injection system for injecting cooling water is provided. And there are a static containment vessel cooling system and a gravity drop type core cooling system as a static emergency core water injection system of a simple boiling water type power generation facility.

単純沸騰水型発電設備の静的格納容器冷却系は、図５に示すように構成されている。炉心５を収納する原子炉圧力容器６に接続される図示しない配管が破壊された場合、蒸気が上部空間である上部空間（ドライウェル）２内に放出される。放出された蒸気は、蒸気入口配管１１を通じてドライウェル２の上部に形成された頂上空間４内のコンデンサ１０に移送されて冷却され、凝縮された水は戻り配管１２を通じてドライウェル２内の重力落下式炉心冷却系のプール７（又は原子炉圧力容器６）に戻している。前記コンデンサ１０には、発生した非凝縮性の気体を下部空間３内のサプレッションプールへ導くベントライン１３を設けている。   The static containment vessel cooling system of the simple boiling water power generation facility is configured as shown in FIG. When a pipe (not shown) connected to the reactor pressure vessel 6 that houses the core 5 is broken, steam is released into the upper space (dry well) 2 that is the upper space. The discharged steam is transferred to the condenser 10 in the top space 4 formed in the upper part of the dry well 2 through the steam inlet pipe 11 and cooled, and the condensed water is dropped by gravity in the dry well 2 through the return pipe 12. It is returned to the pool 7 (or the reactor pressure vessel 6) of the core cooling system. The condenser 10 is provided with a vent line 13 that guides the generated non-condensable gas to the suppression pool in the lower space 3.

このような構成において、静的格納容器冷却系の駆動水頭圧は、ドライウェル２と下部空間３内を連通するベント管１４の下端と前記サプレッションプールの水面間の水位差Δｈ１となる。   In such a configuration, the driving head pressure of the static containment vessel cooling system is a water level difference Δh1 between the lower end of the vent pipe 14 communicating with the dry well 2 and the lower space 3 and the water surface of the suppression pool.

一方、単純沸騰水型発電設備の重力落下式炉心冷却系は、冷却材喪失事故等で原子炉圧力容器６から放出された蒸気を、連通管９を通して重力落下式炉心冷却系のプール７内の空間部に導き、ドライウェル２とプール７内の空間圧力を均圧化させることにより、原子炉圧力容器６とプール７の水位差Δｈ２を利用してプール７が保有する冷却水を、注入配管８を介して原子炉圧力容器６内に注水する構成としている。   On the other hand, the gravity drop type core cooling system of the simple boiling water type power generation facility has the steam released from the reactor pressure vessel 6 due to the loss of coolant accident or the like in the pool 7 of the gravity drop type core cooling system through the communication pipe 9. By guiding the space pressure in the dry well 2 and the pool 7 to the space portion, the cooling water held by the pool 7 using the water level difference Δh2 between the reactor pressure vessel 6 and the pool 7 is injected into the injection pipe. Water is injected into the reactor pressure vessel 6 through 8.

上記従来の技術においては、静的非常炉心注水系の作動時や苛酷事故対策設備として設置された格納容器スプレイ（図示せず）の作動により、ドライウェル２内に放出された蒸気が冷却されて凝縮し、それによって下部空間３内のサプレッションプールの水面上の空間（ウエットウェル３Ａ）に対してドライウェル４内が過度に負圧となって原子炉格納容器１が損傷しないように、ドライウェル２とウエットウェル３Ａ間に真空破壊弁１５を設けている。   In the above conventional technique, the steam discharged into the dry well 2 is cooled by the operation of the containment vessel spray (not shown) installed as a static emergency core water injection system or as a severe accident countermeasure facility. The dry well 4 is condensed so that the inside of the dry well 4 becomes excessively negative with respect to the space above the water surface of the suppression pool in the lower space 3 (wet well 3A) and the reactor containment vessel 1 is not damaged. 2 and a wet well 3A are provided with a vacuum break valve 15.

前記真空破壊弁１５は、図６に示すように、弁体２０とその弁座となるフランジプレート２１とを備え、ドライウェル２がウエットウェル３Ａに対して負圧になったとき、弁対２０が圧力差によって開き、ドライウェル２とウエットウェル３Ａの均圧化を図っている。   As shown in FIG. 6, the vacuum breaker valve 15 includes a valve body 20 and a flange plate 21 serving as a valve seat thereof. When the dry well 2 has a negative pressure with respect to the wet well 3A, the valve pair 20 Is opened by the pressure difference to equalize the pressure in the dry well 2 and the wet well 3A.

このような真空破壊弁１５には、弁の確実なる作動を定例試験時に確認するために、支軸２２に軸支されたアーム２３の一端が弁体２０に連結され、他端がシリンダ２４内を移動するピストン２５に連結されている。そしてシリンダ２４のピストン２５を境とした一方側（図中右側）には、テスト用配管２６を介してテスト用隔離電磁弁２７，テスト用三方電磁弁２８，作動ガス源が接続されている。そして、作動ガス源から例えば圧縮された窒素ガスをテスト用配管２６を介してシリンダ２４内に供給することで、ピストン２５及びアーム２３を介して、弁体２０をフランジプレート２１から引き離して真空破壊弁１５を強制的に開放できるのである。またシリンダ２４のピストン２５を境とした他方側（図中左側）には、配管２９を介して隔離電磁弁３０，三方電磁弁３１，作動ガス源が接続されている。そして、そして、作動ガス源から例えば圧縮された窒素ガスを配管２９を介してシリンダ２４内に供給することで、ピストン２５及びアーム２３を介して、弁体２０をフランジプレート２１に押付けて真空破壊弁１５を強制的に閉じ、苛酷事故時に原子炉圧力容器６及びドライウェル２に冷却水を注水する際に、ドライウェル２内の機密性を高め、設備内保有水の有効利用を図っている。   In such a vacuum breaker valve 15, one end of an arm 23 pivotally supported by a support shaft 22 is connected to the valve body 20 and the other end in the cylinder 24 in order to confirm the reliable operation of the valve during a regular test. It is connected to the piston 25 that moves the. A test isolation solenoid valve 27, a test three-way solenoid valve 28, and a working gas source are connected to one side (right side in the figure) of the cylinder 24 through the piston 25. Then, for example, compressed nitrogen gas is supplied from the working gas source into the cylinder 24 through the test pipe 26, so that the valve body 20 is pulled away from the flange plate 21 through the piston 25 and the arm 23 to break the vacuum. The valve 15 can be forcibly opened. An isolation solenoid valve 30, a three-way solenoid valve 31, and a working gas source are connected to the other side (left side in the figure) of the cylinder 24 with the piston 25 as a boundary through a pipe 29. Then, by supplying, for example, compressed nitrogen gas from the working gas source into the cylinder 24 via the pipe 29, the valve body 20 is pressed against the flange plate 21 via the piston 25 and the arm 23 to break the vacuum. The valve 15 is forcibly closed, and when water is poured into the reactor pressure vessel 6 and the dry well 2 in a severe accident, the confidentiality in the dry well 2 is increased and effective use of the water retained in the facility is intended. .

上記真空破壊弁と同等の技術は、例えば特許文献１に記載されている。   A technique equivalent to the vacuum break valve is described in, for example, Patent Document 1.

特開２００１−１８３４８７号公報JP 2001-183487 A

上記従来における原子炉格納設備は、原子炉液層部から大量の破断流量があると、放出された蒸気の凝縮によりドライウェル２内の圧力が急激に低下し、ウエットウェル３Ａに対して負圧となるので真空破壊弁１５が開放する。また、原子炉格納容器１の格納容器スプレイ（図示せず）等を作動させた場合にも、同様にドライウェル２内の圧力が急激に低下するので真空破壊弁１５が開放する。   In the conventional reactor containment equipment, when there is a large flow rate of breakage from the reactor liquid layer, the pressure in the dry well 2 rapidly decreases due to condensation of the released steam, and a negative pressure is applied to the wet well 3A. Therefore, the vacuum break valve 15 is opened. In addition, when the containment vessel spray (not shown) of the reactor containment vessel 1 is operated, the pressure in the dry well 2 is similarly suddenly lowered, so that the vacuum break valve 15 is opened.

真空破壊弁１５が開放された後も引き続き原子炉圧力容器６から蒸気の放出があると、ドライウェル２と下部空間のウエットウェル３Ａには、ベント管１４の水位差Δｈ１に相当する圧力差が生じる。この状態で、ドライウェル２からウエットウェル３Ａに、真空破壊弁１５を通じた内包気体の漏洩が生じた場合には、ドライウェル２側の蒸気がウエットウェル３Ａ側に流出するので、ウエットウェル３Ａの圧力が上がり水位差Δｈ１が小さくなる。静的格納容器冷却系は原子炉格納容器１の熱除去に際して、コンデンサ１０内に取り込んだ蒸気に含まれる非凝縮性ガスを、前記水位差Δｈ１を利用して下部空間３内のサプレッションプールへ導くように構成されているため、水位差Δｈ１が小さくなると、非凝縮性ガスの排出が行えず、コンデンサ１０内に非凝縮性ガスが蓄積されるので、コンデンサ１０の除熱作用を停止させる問題が生じる。   If steam continues to be released from the reactor pressure vessel 6 after the vacuum breaker valve 15 is opened, a pressure difference corresponding to the water level difference Δh1 of the vent pipe 14 is generated in the dry well 2 and the wet well 3A in the lower space. Arise. In this state, when the inclusion gas leaks from the dry well 2 to the wet well 3A through the vacuum break valve 15, the vapor on the dry well 2 side flows out to the wet well 3A side. The pressure increases and the water level difference Δh1 decreases. The static containment vessel cooling system guides the non-condensable gas contained in the steam taken into the condenser 10 to the suppression pool in the lower space 3 using the water level difference Δh1 when removing heat from the reactor containment vessel 1. Therefore, when the water level difference Δh1 becomes small, the non-condensable gas cannot be discharged, and the non-condensable gas is accumulated in the condenser 10, so that there is a problem of stopping the heat removal action of the condenser 10. Arise.

本発明の目的は、ドライウェル（上部空間）側からウエットウェル（下部空間）側への蒸気の漏洩を防止し、コンデンサの除熱作用を継続できる原子炉格納設備を提供することにある。   An object of the present invention is to provide a reactor containment facility capable of preventing the leakage of steam from the dry well (upper space) side to the wet well (lower space) side and continuing the heat removal action of the condenser.

本発明は上記目的を達成するために、原子炉圧力容器が収容された上部空間と、サプレッションプールを形成した下部空間と、前記上部空間の圧力が前記下部空間の圧力よりも低くなったときに開放する真空破壊弁と、非常時に前記上部空間内に放出された蒸気を回収して除熱し凝縮された水を前記上部空間内に戻し非凝縮性ガスを前記下部空間へ導くコンデンサと、非常時に前記上部空間内に冷却水の注入を行う非常炉心注水系とを備えた原子炉格納設備において、前記非常炉心注水系を起動させる注水系起動手段と、前記上部空間と前記下部空間との差圧を検出する圧力検出手段とを備え、前記注水系起動手段からの起動信号と前記圧力検出手段からの前記上部空間の圧力が前記下部空間の圧力よりも高くなったことを示す信号とを条件に前記真空破壊弁を強制的に閉弁する閉弁機構を設けたのである。   In order to achieve the above object, the present invention provides an upper space in which a reactor pressure vessel is accommodated, a lower space in which a suppression pool is formed, and a pressure in the upper space that is lower than a pressure in the lower space. A vacuum breaker valve that is opened, a condenser that recovers the steam released into the upper space in the event of an emergency, removes the condensed water and returns the condensed water to the upper space, and a non-condensable gas to the lower space; In a nuclear reactor containment facility equipped with an emergency core water injection system for injecting cooling water into the upper space, water injection system starting means for starting the emergency core water injection system, and a differential pressure between the upper space and the lower space And a pressure detection means for detecting a condition of a start signal from the water injection system start means and a signal indicating that the pressure in the upper space from the pressure detection means is higher than the pressure in the lower space. Forcing it had provided a valve closing mechanism for closing the vacuum breaker valve.

上記のように注水系起動手段からの起動信号と上部空間の圧力が下部空間の圧力よりも高くなったことを示す信号とを条件に真空破壊弁を強制的に閉弁することで、注水起動手段の作動時における上部空間から下部空間への上記の流出を防止することができ、これによって上部空間から下部空間の間に発生した圧力差を利用しコンデンサ内の非凝縮性ガスを下部空間へ導くことができ、コンデンサの除熱作用を促進することができるのである。   As described above, water injection start is performed by forcibly closing the vacuum breaker valve on the condition of the start signal from the water injection system starting means and the signal indicating that the pressure in the upper space is higher than the pressure in the lower space. It is possible to prevent the above-mentioned outflow from the upper space to the lower space during operation of the means, thereby utilizing the pressure difference generated between the upper space and the lower space to transfer the non-condensable gas in the condenser to the lower space. Therefore, the heat removal action of the capacitor can be promoted.

以上説明したように本発明によれば、ドライウェル（上部空間）側からウエットウェル（下部空間）側への蒸気の漏洩を防止し、コンデンサの除熱作用を継続できる原子炉格納設備を得ることができる。   As described above, according to the present invention, it is possible to obtain a reactor containment facility capable of preventing the leakage of steam from the dry well (upper space) side to the wet well (lower space) side and continuing the heat removal action of the capacitor. Can do.

以下本発明による原子炉格納設備を図１に基づいて説明する。尚、図５と同一符号は、同一部品を示すので、再度の詳細な説明は省略する。   A reactor containment facility according to the present invention will be described below with reference to FIG. Note that the same reference numerals as those in FIG. 5 indicate the same parts, and thus detailed description thereof is omitted.

原子炉格納容器１は、上部空間（ドライウェル）２と、その下側の下部空間３と、上部空間２の上方に形成された頂部空間４とを有している。   The reactor containment vessel 1 has an upper space (dry well) 2, a lower space 3 below the upper space 2, and a top space 4 formed above the upper space 2.

ドライウェル２内には、炉心５を収容した原子炉圧力容器６と、この格納容器６よりも上方に配置された重力落下式炉心冷却系のプール７が設置されており、プール７からは注入配管８を介して原子炉圧力容器６内に注水されるように構成されている。また、プール７と原子炉圧力容器６とは、原子炉圧力容器６内と同圧となるように、連通管９で連通している。   In the dry well 2, a reactor pressure vessel 6 containing a core 5 and a gravity drop type core cooling system pool 7 disposed above the containment vessel 6 are installed. Water is poured into the reactor pressure vessel 6 through the pipe 8. Further, the pool 7 and the reactor pressure vessel 6 are communicated with each other through a communication pipe 9 so as to have the same pressure as that in the reactor pressure vessel 6.

頂部空間４内には、コンデンサ１０が設置されており、コンデンサ１０には、ドライウェル２に開口する蒸気入口配管１１が接続されており、また、凝縮水を前記プール７に戻す戻り配管１２が接続されている。さらに、コンデンサ１０からは、下部空間３に延在するベントライン１３が接続されている。   A condenser 10 is installed in the top space 4, a steam inlet pipe 11 that opens to the dry well 2 is connected to the condenser 10, and a return pipe 12 that returns condensed water to the pool 7. It is connected. Further, a vent line 13 extending to the lower space 3 is connected from the capacitor 10.

前記ドライウェル２の下部からは、先端が前記下部空間３内のサプレッションプールに至る連通するベント管１４が接続されている。そして、前記ドライウェル２と前記下部空間３のサプレッションプールの水面上の空間であるウエットウェル３Ａとは、真空破壊弁１５を介して連通している。   A vent pipe 14 is connected from the lower portion of the dry well 2 to a tip communicating with the suppression pool in the lower space 3. The dry well 2 and the wet well 3 </ b> A, which is a space above the surface of the suppression pool in the lower space 3, communicate with each other via a vacuum break valve 15.

このほか、前記ドライウェル２内の圧力を検出する圧力計１６と、ウエットウェル３Ａ内の圧力を検出する圧力計１７と、これら圧力計１６，１７の圧力を比較する差圧計１８とが設けられており、前記真空破壊弁１５は、前記差圧計１８の計測値が所定値を超えた場合で、強制的に閉弁するように構成されている。さらに、前記真空破壊弁１５は、前記差圧計１８の計測値に基づいて閉弁するほか、冷却材喪失事故時に原子炉格納容器の冷却及び炉心へ冷却水を注水する静的非常炉心注水系の作動を条件に閉弁するように構成されている。尚、前記静的非常炉心注水系の構成及び動作は、既述の静的格納容器冷却系と重力落下式炉心冷却系及びの通りであり、原子炉格納設備の動作も基本的に既述の通りである。   In addition, a pressure gauge 16 for detecting the pressure in the dry well 2, a pressure gauge 17 for detecting the pressure in the wet well 3A, and a differential pressure gauge 18 for comparing the pressures of the pressure gauges 16 and 17 are provided. The vacuum breaker valve 15 is configured to forcibly close the valve when the measured value of the differential pressure gauge 18 exceeds a predetermined value. Further, the vacuum breaker valve 15 is closed based on the measured value of the differential pressure gauge 18, and in addition to a static emergency core injection system that cools the reactor containment vessel and injects coolant into the core in the event of a coolant loss accident. The valve is configured to close on the condition of operation. The configuration and operation of the static emergency core water injection system are the same as those described above for the static containment vessel cooling system and the gravity drop type core cooling system, and the operation of the reactor containment equipment is also basically as described above. Street.

また、真空破壊弁１５は、図２に示すように、基本構成は、図６に示す従来の真空破壊弁１５と同じ構成であり、したがって、図６と同一符号は、同一部品を示すので、再度の詳細な説明は省略する。しかしながら、真空破壊弁１５を強制的に開閉するためにテスト用隔離電磁弁２７，テスト用三方電磁弁２８及び隔離電磁弁３０，三方電磁弁３１を駆動する信号が従来と大きく異なる。   Further, as shown in FIG. 2, the basic configuration of the vacuum breaking valve 15 is the same as that of the conventional vacuum breaking valve 15 shown in FIG. 6, and therefore the same reference numerals as those in FIG. The detailed description again will be omitted. However, the signals for driving the test isolation solenoid valve 27, the test three-way solenoid valve 28, the isolation solenoid valve 30, and the three-way solenoid valve 31 in order to forcibly open and close the vacuum breaker valve 15 are significantly different from the conventional ones.

即ち、ドライウェル２内の圧力からウエットウェル３Ａ内の圧力を差し引いた圧力が正圧、云い代えればドライウェル２内の圧力がウエットウェル３Ａ内の圧力より高い場合に真空破壊弁１５を強制的に閉弁させる強制閉信号を隔離電磁弁３０，三方電磁弁３１に送信して駆動する。この時の閉弁操作は、差圧計１８での計測値が、ドライウェル２内の圧力からウエットウェル３Ａ内の圧力を差し引いた圧力が正圧の場合、即ち、ウェル間差圧が正の場合、図３に示すように、静的非常炉心注水系である静的格納容器冷却系の動作信号（ＥＣＣＳ起動信号）の発生と共に、ウェル間差圧が正の信号の発生により、真空破壊弁１５の強制閉信号が発せられ、この強制閉信号により隔離電磁弁３０，三方電磁弁３１を開いて圧縮機体である窒素ガスをシリンダ２４内に供給し、弁体２０をフランジプレート２１に密着させることで真空破壊弁１５は強制的に閉弁される。したがって、真空破壊弁１５が開放された後も引き続き原子炉圧力容器６から蒸気の放出があって、ドライウェル２と下部空３間のウエットウェル３Ａに、ベント管１４の水位差Δｈ１に相当する圧力差が生じても真空破壊弁１５を通じた内包気体の漏洩が生じることはなく、その結果、必要な水位差Δｈ１を維持できるので、コンデンサ１０内に取り込んだ蒸気に含まれる非凝縮性ガスを、この水位差Δｈ１を利用して下部空間３内のサプレッションプールへ導くことができ、コンデンサ１０の除熱作用を維持することができるのである。   That is, when the pressure obtained by subtracting the pressure in the wet well 3A from the pressure in the dry well 2 is a positive pressure, in other words, the pressure in the dry well 2 is higher than the pressure in the wet well 3A, the vacuum break valve 15 is forced. The forcible closing signal to be closed is transmitted to the isolation solenoid valve 30 and the three-way solenoid valve 31 for driving. The valve closing operation at this time is performed when the pressure measured by the differential pressure gauge 18 is a positive pressure obtained by subtracting the pressure in the wet well 3A from the pressure in the dry well 2, that is, the pressure difference between the wells is positive. As shown in FIG. 3, when the operation signal (ECCS activation signal) of the static containment vessel cooling system, which is a static emergency core water injection system, is generated, a signal with a positive pressure difference between the wells is generated. The forcible closing signal is issued, the isolation electromagnetic valve 30 and the three-way electromagnetic valve 31 are opened by this forcible closing signal, nitrogen gas as a compressor body is supplied into the cylinder 24, and the valve body 20 is brought into close contact with the flange plate 21. Thus, the vacuum breaker valve 15 is forcibly closed. Therefore, even after the vacuum breaker valve 15 is opened, steam is continuously released from the reactor pressure vessel 6, which corresponds to the water level difference Δh 1 of the vent pipe 14 in the wet well 3 A between the dry well 2 and the lower space 3. Even if a pressure difference occurs, leakage of the encapsulated gas through the vacuum break valve 15 does not occur. As a result, the necessary water level difference Δh1 can be maintained, so that the non-condensable gas contained in the steam taken into the condenser 10 can be maintained. The water level difference Δh1 can be used to lead to the suppression pool in the lower space 3, and the heat removal action of the condenser 10 can be maintained.

また、ウェル間差圧が負、云い代えればドライウェル２の圧力がウエットウェル３Ａよりも低圧となった場合には、ウェル間差圧が正の信号の発生がなく、その結果、真空破壊弁１５の強制閉が解除されることから、ウェル間差圧により真空破壊弁１５は開弁する。尚、図示しない手動スイッチによるテスト信号により、テスト用隔離電磁弁２７，テスト用三方電磁弁２８を開いて真空破壊弁１５を強制開している最中に、真空破壊弁１５を強制閉する信号（ＥＣＣＳ起動信号とウェル間差圧が正の信号）が発生した場合には、テスト時の強制開信号を強制的にワイプアウトして、ＥＣＣＳ起動信号とウェル間差圧が正の信号を優先する。   Further, when the pressure difference between the wells is negative, in other words, when the pressure of the dry well 2 is lower than that of the wet well 3A, there is no generation of a signal with a positive pressure difference between the wells. Since the forcible closing of 15 is released, the vacuum breaker valve 15 is opened by the inter-well differential pressure. A signal for forcibly closing the vacuum breaker valve 15 while the test isolation solenoid valve 27 and the test three-way solenoid valve 28 are opened and the vacuum breaker valve 15 is forcibly opened by a test signal from a manual switch (not shown). If (ECCS activation signal and well differential pressure is positive signal) occurs, forced open signal during test is forcibly wiped out, and ECCS activation signal and positive differential pressure between wells have priority. To do.

ところで、前記真空破壊弁１５を強制的に閉弁させる代わりに、真空破壊弁１５が開閉する通路に、別の電磁弁を設け、この電磁弁を緊急時に閉弁するようにしてもよい。   By the way, instead of forcibly closing the vacuum breaker valve 15, another solenoid valve may be provided in the passage where the vacuum breaker valve 15 opens and closes, and this solenoid valve may be closed in an emergency.

図４は、静的格納容器冷却系の性能維持が要求される時間が、事故発生直後ではなく、事故発生後から所定時間経過後内に要求される場合の実施の形態を示すもので、図３に示す強制閉信号の条件としてのＥＣＣＳ起動信号とウェル間差圧が正の信号に、運転員による強制閉スイッチ操作による強制閉信号を加えたことを条件に真空破壊弁１５の強制閉信号を発生させるようにしたものである。   FIG. 4 shows an embodiment in which the time required to maintain the performance of the static containment cooling system is required not after the occurrence of the accident but within a predetermined time after the occurrence of the accident. The forced closing signal of the vacuum breaker valve 15 is provided on the condition that the forced closing signal by the forced closing switch operation by the operator is added to the ECCS activation signal and the positive differential pressure between wells as the forced closing signal condition shown in FIG. Is generated.

本発明による原子炉格納設備の一実施の形態を示すブロック図。The block diagram which shows one Embodiment of the nuclear reactor containment equipment by this invention. 図１に用いられる真空破壊弁の構成を示すブロック図。The block diagram which shows the structure of the vacuum breaker valve used for FIG. 図２の真空破壊弁の開閉動作を指示する信号系を示すブロック図。The block diagram which shows the signal system which instruct | indicates the opening / closing operation | movement of the vacuum breaker valve of FIG. 図３の信号系の変形例を示すブロック図。The block diagram which shows the modification of the signal system of FIG. 従来による原子炉格納設備を示すブロック図。The block diagram which shows the nuclear reactor containment equipment by the former. 図５に用いられる真空破壊弁の構成を示すブロック図。The block diagram which shows the structure of the vacuum breaker valve used for FIG.

符号の説明Explanation of symbols

１…原子炉格納容器、２…上部空間（ドライウェル）、３…下部空間、３Ａ…ウエットウェル、４…頂部空間、５…炉心、６…原子炉圧力容器、７…プール、８…注入配管、９…連通管、１０…コンデンサ、１１…蒸気入口配管、１２…戻り配管、１３…ベントライン、１４…ベント管、１５…真空破壊弁、１６，１７…圧力計、１８…差圧計、２０…弁体、２１…フランジプレート、２３…アーム、２４…シリンダ、２５…ピストン、２７…テスト用隔離電磁弁、２８…テスト用三方電磁弁、３０…隔離電磁弁、３１…三方電磁弁。   DESCRIPTION OF SYMBOLS 1 ... Reactor containment vessel, 2 ... Upper space (dry well), 3 ... Lower space, 3A ... Wet well, 4 ... Top space, 5 ... Reactor core, 6 ... Reactor pressure vessel, 7 ... Pool, 8 ... Injection piping , 9 ... Communication pipe, 10 ... Condenser, 11 ... Steam inlet pipe, 12 ... Return pipe, 13 ... Vent line, 14 ... Vent pipe, 15 ... Vacuum break valve, 16, 17 ... Pressure gauge, 18 ... Differential pressure gauge, 20 DESCRIPTION OF SYMBOLS ... Valve body, 21 ... Flange plate, 23 ... Arm, 24 ... Cylinder, 25 ... Piston, 27 ... Test isolation solenoid valve, 28 ... Test three-way solenoid valve, 30 ... Isolation solenoid valve, 31 ... Three-way solenoid valve

Claims (3)

原子炉圧力容器が収容された上部空間と、サプレッションプールを形成した下部空間と、前記上部空間の圧力が前記下部空間の圧力よりも低くなったときに開放する真空破壊弁と、非常時に前記上部空間内に放出された蒸気を回収し非凝縮性ガスを前記下部空間へ導くコンデンサと、非常時に前記原子炉圧力容器内に冷却水の注入を行う非常炉心注水系とを備えた原子炉格納設備において、前記非常炉心注水系を起動させる注水系起動手段と、前記上部空間と前記下部空間との差圧を検出する圧力検出手段とを備え、前記注水系起動手段からの起動信号と前記圧力検出手段からの前記上部空間の圧力が前記下部空間の圧力よりも高くなったことを示す信号とを条件に前記真空破壊弁を強制的に閉弁する閉弁機構を設けたことを特徴とする原子炉格納設備。   An upper space in which a reactor pressure vessel is accommodated, a lower space in which a suppression pool is formed, a vacuum breaker valve that opens when the pressure in the upper space becomes lower than the pressure in the lower space, and the upper space in an emergency Reactor containment facility comprising a condenser that collects steam discharged into the space and guides noncondensable gas to the lower space, and an emergency core injection system that injects cooling water into the reactor pressure vessel in an emergency. A starting water injection system for starting the emergency core water injection system, and a pressure detection means for detecting a differential pressure between the upper space and the lower space, the start signal from the injection system and the pressure detection And a valve closing mechanism for forcibly closing the vacuum breaker valve on condition that a signal indicating that the pressure in the upper space is higher than the pressure in the lower space is provided. Storage facility. 前記真空破壊弁を強制的に閉弁する閉弁機構は、前記注水系起動手段からの起動信号と前記圧力検出手段からの前記上部空間の圧力が前記下部空間の圧力よりも高くなったことを示す信号及び手動操作による強制閉信号とを条件に閉弁動作を行うように構成されていることを特徴とする請求項１記載の原子炉格納設備。   The valve closing mechanism for forcibly closing the vacuum breaker valve is that the start signal from the water injection system starting means and the pressure in the upper space from the pressure detecting means are higher than the pressure in the lower space. The reactor containment equipment according to claim 1, wherein a valve closing operation is performed under a condition of a signal to be indicated and a forced closing signal by manual operation. 原子炉圧力容器が収容された上部空間と、サプレッションプールを形成した下部空間と、前記上部空間の圧力が前記下部空間の圧力よりも低くなったときに開放する真空破壊弁と、非常時に前記原子炉圧力容器内に冷却水の注入を行う非常炉心注水系とを備えた原子炉格納設備の圧力制御方法において、前記非常炉心注水系の起動時と前記上部空間の圧力が前記下部空間の圧力よりも高くなった時とが重なった時に前記真空破壊弁を強制的に閉じるようにした原子炉格納設備の圧力制御方法。 An upper space in which a reactor pressure vessel is accommodated, a lower space in which a suppression pool is formed, a vacuum breaker valve that is opened when the pressure in the upper space becomes lower than the pressure in the lower space, and the atoms in an emergency. In a pressure control method for a containment facility equipped with an emergency core water injection system for injecting cooling water into a reactor pressure vessel, the pressure in the upper space and the pressure in the lower space are more A pressure control method for a containment of a nuclear reactor, in which the vacuum breaker valve is forcibly closed when it becomes higher.
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JPH06194484A (en) * 1992-12-24 1994-07-15 Toshiba Corp Nuclear reactor containment vessel cooling facility
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JP2001183487A (en) * 1999-12-24 2001-07-06 Hitachi Ltd Water filling facilities for reactor jessel

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JPH06194484A (en) * 1992-12-24 1994-07-15 Toshiba Corp Nuclear reactor containment vessel cooling facility
JPH06222181A (en) * 1993-01-22 1994-08-12 Toshiba Corp Cooling facility of nuclear reactor containment
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