JPS61228391A - Emergency core cooling method and device - Google Patents

Emergency core cooling method and device

Info

Publication number
JPS61228391A
JPS61228391A JP60069423A JP6942385A JPS61228391A JP S61228391 A JPS61228391 A JP S61228391A JP 60069423 A JP60069423 A JP 60069423A JP 6942385 A JP6942385 A JP 6942385A JP S61228391 A JPS61228391 A JP S61228391A
Authority
JP
Japan
Prior art keywords
temperature
cooling water
flow rate
primary coolant
pressure
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
Application number
JP60069423A
Other languages
Japanese (ja)
Other versions
JPH0713669B2 (en
Inventor
道雄 村瀬
良之 片岡
孝志 池田
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.)
Hitachi Ltd
Original Assignee
Hitachi 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 Hitachi Ltd filed Critical Hitachi Ltd
Priority to JP60069423A priority Critical patent/JPH0713669B2/en
Publication of JPS61228391A publication Critical patent/JPS61228391A/en
Publication of JPH0713669B2 publication Critical patent/JPH0713669B2/en
Anticipated expiration legal-status Critical
Expired - Lifetime legal-status Critical Current

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

  • Heat Treatments In General, Especially Conveying And Cooling (AREA)
  • Casting Devices For Molds (AREA)
  • Blast Furnaces (AREA)
  • Control Of Temperature (AREA)

Abstract

(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

【発明の詳細な説明】 〔発明の利用分野〕 本発明は、原子炉の非常用炉心冷却方法とその装置に係
り、特に低温水の注入で発生する熱衝撃を防止するのに
好適な非常用炉心冷却方法及び装置に関する。[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to an emergency core cooling method and device for a nuclear reactor, and particularly to an emergency core cooling method suitable for preventing thermal shock caused by low-temperature water injection. This invention relates to core cooling methods and devices.

〔発明の背景〕[Background of the invention]

原子炉には、事故時に炉心を冷却するために非常用炉心
冷却装置が設けられており、例えば原子カニ業、第１０
巻、第１号における医用による「原子炉圧力容器の健全
性に関する最近の話題」と題する文献に示されているよ
うに、原子炉容器内が高温、高圧状態である時に非常用
炉心冷却装置で低温水を注入した場合、原子炉容器の急
冷による熱衝撃で原子炉容器が損傷を受ける可能性が指
摘されている。また、前記文献で論じられているように
、熱衝撃の回避策として、（１）非常用炉心冷却装置で
供給される冷却水の昇温、（２）放射線照射による原子
炉容器の材料劣化の回復を目的とした原子炉容器の焼な
まし、（３）原子炉容器の内面と外面の温度差を小さく
することを目的とした原子炉容器外面の冷却、及び（４
）ＪＪＫ子炉容器が受ける放射線照射量を低減化するた
めの放射線の反射体などが検討されているが、いずれも
過大な設備や費用を必要とし具体化されていない。Nuclear reactors are equipped with emergency core cooling systems to cool the core in the event of an accident.
As shown in the medical article entitled "Recent Topics on the Integrity of Reactor Pressure Vessels" in Volume 1, No. 1, the emergency core cooling system is It has been pointed out that if low-temperature water is injected, the reactor vessel may be damaged by thermal shock caused by rapid cooling of the reactor vessel. In addition, as discussed in the above literature, measures to avoid thermal shock include (1) increasing the temperature of the cooling water supplied by the emergency core cooling system, and (2) preventing material deterioration of the reactor vessel due to radiation irradiation. Annealing of the reactor vessel for the purpose of recovery, (3) cooling of the outer surface of the reactor vessel for the purpose of reducing the temperature difference between the inner and outer surfaces of the reactor vessel, and (4)
) Radiation reflectors are being considered to reduce the amount of radiation irradiated to the JJK reactor vessel, but none of these have been implemented as they require excessive equipment and costs.

一方、非常用炉心冷却装置で供給される冷却水を昇温す
る先行技術例としては、特開昭５３−５１３９５及び特
公昭５１−３１３９５が知られている。前者の公知例は
、信頼性向上を目的として原子炉容器内に冷却材貯蔵タ
ンクを設けたもので、冷却材貯蔵量が著しく制限される
欠点がある。後者の公知例は、炉心冷却性能の向上を目
的として冷却水貯蔵水槽からの冷却水と高温の一次冷却
水とをポンプの上流部で混合することにより冷却水の温
度を５５℃〜１００℃の範囲で加熱するもので、冷却水
貯水槽内の圧力に対する飽和温度以上に加熱できない欠
点がある（飽和温度以上に加熱すると冷却水が沸騰し、
ポンプが損傷する危険があり、容器の耐圧上の問題もあ
る）。On the other hand, as prior art examples of raising the temperature of cooling water supplied by an emergency core cooling system, Japanese Patent Laid-Open No. 53-51395 and Japanese Patent Publication No. 51-31395 are known. The former known example is one in which a coolant storage tank is provided within the reactor vessel for the purpose of improving reliability, and has the disadvantage that the amount of coolant stored is severely limited. The latter known example mixes cooling water from a cooling water storage tank and high-temperature primary cooling water at the upstream part of a pump to increase the temperature of the cooling water to 55°C to 100°C for the purpose of improving core cooling performance. It heats within a range, and has the disadvantage that it cannot be heated above the saturation temperature for the pressure in the cooling water storage tank (if heated above the saturation temperature, the cooling water will boil,
There is a risk of damage to the pump, and there is also a problem with the pressure resistance of the container).

〔発明の目的〕[Purpose of the invention]

本発明の目的は、簡単かつ小模規な構造で多量の冷却水
を冷却水貯水槽内の圧力に対する飽和温度以上に加熱で
き、原子炉容器に生ずる熱衝撃の防止が可能な非常用炉
心冷却方法とその装置を提供することである。An object of the present invention is to provide emergency core cooling that can heat a large amount of cooling water to a temperature higher than the saturation temperature for the pressure in the cooling water storage tank with a simple and small-scale structure, and that can prevent thermal shock occurring in the reactor vessel. An object of the present invention is to provide a method and an apparatus therefor.

（発明の概要〕 本発明は、冷却水貯水槽からの冷却水と高温の一次冷却
材とを冷却水の給水ポンプの下流部で混合し、混合水の
温度を検出して一次冷却材及び冷却水の流量を調整する
ことによって混合水の温度を冷却水貯水槽内の圧力に対
する飽和温度より高くした後、原子炉容器に供給し、熱
衝撃による原子炉容器の損傷を防止するものである。(Summary of the Invention) The present invention mixes cooling water from a cooling water storage tank and a high-temperature primary coolant at a downstream part of a cooling water supply pump, detects the temperature of the mixed water, and then mixes the cooling water from a cooling water storage tank with a high-temperature primary coolant. By adjusting the water flow rate, the temperature of the mixed water is made higher than the saturation temperature for the pressure in the cooling water storage tank, and then it is supplied to the reactor vessel to prevent damage to the reactor vessel due to thermal shock.

以下、本発明に至るまでの技術的困難性と本発明の原理
について説明する。Hereinafter, the technical difficulties leading up to the present invention and the principle of the present invention will be explained.

通常運転圧力が高く、低温の冷却水の注入によって熱衝
撃が発生しやすい加圧水型原子炉を用いて本発明に至る
までの技術的困難性について説明する。The technical difficulties leading up to the present invention will be explained using a pressurized water reactor, which normally has a high operating pressure and is prone to thermal shock due to the injection of low-temperature cooling water.

第２図は従来技術による加圧木型原子炉と非常用炉心冷
却装置及び冷却水の加熱方法を示す０通常運転時には、
原子炉容器１内の一次冷却材４は炉心２で加熱され、ホ
ットレグ５を通って蒸気発生器６内の伝熱管７で冷却さ
れた後、ｍ環ポンプ８で駆動されてコールドレグ９を通
り原子炉容器１内のダウンカマ３から炉心２に再循環さ
れる。Figure 2 shows a pressurized wooden nuclear reactor, an emergency core cooling system, and a cooling water heating method according to the prior art.During normal operation,
The primary coolant 4 in the reactor vessel 1 is heated in the reactor core 2, passes through the hot leg 5, is cooled in the heat exchanger tube 7 in the steam generator 6, is driven by the m-ring pump 8, passes through the cold leg 9, and is the atomic It is recirculated from the downcomer 3 in the reactor vessel 1 to the reactor core 2.

一方、二次冷却材１５は蒸気発生器６の伝熱管７で加熱
され沸騰し、発生した蒸気でタービンを駆動し発電に供
せられる。このようなホットレグ５゜蒸気発生器６．循
環ポンプ８．及びコールドレグ９から成るループは複数
系統設置されており、一系統には原子炉容器１内の圧力
を制御するための加圧器１０が設置されている。On the other hand, the secondary coolant 15 is heated and boiled in the heat exchanger tube 7 of the steam generator 6, and the generated steam drives a turbine and is used for power generation. Hot leg 5° steam generator like this6. Circulation pump 8. A plurality of loops consisting of the cold legs 9 and 9 are installed, and one system is installed with a pressurizer 10 for controlling the pressure inside the reactor vessel 1.

万一、ホットレグ５もしくはコールドレグ９などの一次
冷却材配管が破断し一次冷却材４が流出するような冷却
材喪失事故が発生した場合でも、炉心２を冷却できるよ
うに複数系統の非常用炉心冷却装置が設置されている。In the unlikely event that a coolant loss accident occurs such as a primary coolant piping such as hot leg 5 or cold leg 9 rupturing and primary coolant 4 flowing out, multiple emergency core cooling systems are installed so that the core 2 can be cooled. Equipment is installed.

第２図に示した非常用炉心冷却装置は、多量の冷却水２
２を貯水した冷却水貯水槽２１と給水ポンプ２３で構成
されており、事故時には逆止弁２４及び流量調節弁２５
を介して冷却水２２がコールドレグ９もしくはホットレ
グ５を通って炉心２に供給される。この時、放射線の照
射で原子炉容器１の材料強度が劣化した状態で（放射線
の照射により容器壁材料が延性を失なう温度が２００℃
程度に上昇しているといわれている）、かつ、原子炉容
器１内の圧力が高い時に（圧力が２０気圧未満では、後
述する内圧によるフープ応力が十分小さくなり容器壁材
料が損傷する危険性はないといわれている）、低温の冷
却水が注入されると、［子炉容器１は延性を失なう２０
０℃以下に冷却され、脆性特性のみを示し強度的に弱く
なると共に熱衝撃による熱応力と内圧によるフープ応力
によって原子炉容器１の溶接部が損傷を受ける可能性が
あることが指摘されている。The emergency core cooling system shown in Figure 2 uses a large amount of cooling water 2.
The system consists of a cooling water storage tank 21 storing water of
Cooling water 22 is supplied to the core 2 through the cold leg 9 or the hot leg 5. At this time, the material strength of the reactor vessel 1 has deteriorated due to radiation irradiation (the temperature at which the vessel wall material loses ductility due to radiation irradiation is 200°C).
), and when the pressure inside the reactor vessel 1 is high (if the pressure is less than 20 atmospheres, the hoop stress due to the internal pressure, which will be described later, will be sufficiently small and there is a risk of damage to the vessel wall material). When low temperature cooling water is injected, the child reactor vessel 1 loses its ductility.
It has been pointed out that when cooled to below 0°C, it exhibits only brittle characteristics and becomes weak in strength, and the welded parts of the reactor vessel 1 may be damaged by thermal stress due to thermal shock and hoop stress due to internal pressure. .

この熱衝撃による損傷を回避するには、内圧によるフー
プ応力が大きくなる２０気圧以上において、容器壁材料
が延性を失なわない２００℃以上に冷却水２２を加熱し
、容器壁の急冷却及び強度低下を回避すればよい。最も
単純な冷却水２２の加熱方法としては、冷却水貯水槽２
１に加熱器Ａを設ければよいが、冷却水貯水槽２１は数
千から数万立方米の容積を有し大きな加熱源を必要とし
。In order to avoid damage caused by this thermal shock, the cooling water 22 is heated to 200 degrees Celsius or higher, at which the container wall material does not lose its ductility, at a temperature of 20 atmospheres or higher, where the hoop stress due to internal pressure becomes large, and the container wall is rapidly cooled and strengthened. All you have to do is avoid the decline. The simplest method for heating the cooling water 22 is to heat the cooling water in the cooling water storage tank 2.
However, the cooling water storage tank 21 has a capacity of several thousand to tens of thousands of cubic meters and requires a large heating source.

かつ、熱衝撃を防止しうる温度（２００℃）以上にまで
冷却水２２を加熱するには冷却水貯水槽２１を耐圧容器
（２００℃に対する水の飽和圧力は約１６気圧）にする
必要があり、非現実的である。給水ポンプ２３の下流部
に加熱器Ｂを設は原子炉容器１に注入する冷却水２２の
みを加熱する方法は現実的であるが、約１０ＭＷの大型
加熱器が必要であり、大幅なコスト上昇となる。In addition, in order to heat the cooling water 22 to a temperature above which thermal shock can be prevented (200°C), the cooling water storage tank 21 must be made into a pressure-resistant container (the saturation pressure of water at 200°C is approximately 16 atmospheres). , unrealistic. Although it is practical to install a heater B downstream of the feed water pump 23 to heat only the cooling water 22 injected into the reactor vessel 1, it requires a large heater of approximately 10 MW, which significantly increases costs. becomes.

一方、原子炉容器１内には高温の一次冷却材４があり、
この一次冷却材４を冷却水２２の加熱源として利用する
ことが可能である。この場合、一次冷却材４と冷却水２
２を直接混合するのが最も効率的である。しかし、特公
昭５１−３１３９５による公知例のように、給水ポンプ
２３の上流部において一次冷却材４と冷却水２２とを混
合するルートＣを採用した場合、加熱器Ａを用いる場合
と同様に、冷却水貯水槽２１内の圧力に対する飽和温度
以上に冷却水２２を加熱できず、熱衝撃の発生を防止で
きない。冷却水貯水槽２１内め圧力に対する飽和温度以
上に冷却水２２を加熱すると給水ポンプ２３内で冷却水
２２が沸騰し、給水ポンプ２３が損傷する危険性がある
。On the other hand, there is a high temperature primary coolant 4 inside the reactor vessel 1.
This primary coolant 4 can be used as a heating source for the cooling water 22. In this case, primary coolant 4 and cooling water 2
It is most efficient to mix 2 directly. However, when route C is adopted in which the primary coolant 4 and the cooling water 22 are mixed in the upstream part of the water supply pump 23, as in the known example published in Japanese Patent Publication No. 51-31395, similar to the case where heater A is used, The cooling water 22 cannot be heated above the saturation temperature with respect to the pressure in the cooling water storage tank 21, and the occurrence of thermal shock cannot be prevented. If the cooling water 22 is heated above the saturation temperature with respect to the internal pressure of the cooling water storage tank 21, the cooling water 22 will boil within the water supply pump 23, and there is a risk that the water supply pump 23 will be damaged.

換言すれば、一次冷却材４を冷却水２２の加熱源に使用
する場合、給水ポンプ２３の下流部で混合するルートＤ
としなければならない、この場合には、給水ポンプ２３
の下流部では給水ポンプ２３の吐出圧力によって加圧さ
れているため、冷却水２２を冷却水貯水槽２１内の圧力
に対する飽和温度より高い任意の温度にまで加熱するこ
とが可能となる。ただし、この場合、前述したように、
給水ポンプ２３の下流部ではすでに加圧されているため
、一次冷却材４を給水ポンプ２３の下流部に供給する手
段（すなわち駆動源）が必要である。In other words, when the primary coolant 4 is used as a heating source for the cooling water 22, the route D is mixed downstream of the water supply pump 23.
In this case, the water supply pump 23
Since the downstream part of the cooling water 22 is pressurized by the discharge pressure of the water supply pump 23, it is possible to heat the cooling water 22 to an arbitrary temperature higher than the saturation temperature with respect to the pressure in the cooling water storage tank 21. However, in this case, as mentioned above,
Since the downstream part of the water supply pump 23 is already pressurized, a means (that is, a driving source) for supplying the primary coolant 4 to the downstream part of the water supply pump 23 is required.

この一次冷却材４の駆動源にはポンプもしくはエゼクタ
などを使用できる。すなわち、ルートＤにより給水ポン
プ２３の下流部において一次冷却材４と冷却水２２を混
合すれば、ルートＤの高温配管とポンプもしくはエゼク
タの追加のみで、冷却水２２を冷却水貯水槽２１内の圧
力に対する飽和温度より高い任意の温度にまで加熱する
ことが可能となる。さらに、この場合、一次冷却材４を
原子炉容器１から給水ポンプ２３の下流部を通って原子
炉容器１に再循環させるため、炉心２の冷却性能が向上
する。しかし、この場合も次のような技術的問題がある
。A pump or an ejector can be used as a driving source for this primary coolant 4. That is, if the primary coolant 4 and the cooling water 22 are mixed downstream of the water supply pump 23 via route D, the cooling water 22 can be transferred to the cooling water storage tank 21 by simply adding the high-temperature piping and pump or ejector of route D. It becomes possible to heat to any temperature higher than the saturation temperature for pressure. Furthermore, in this case, since the primary coolant 4 is recirculated from the reactor vessel 1 to the reactor vessel 1 through the downstream part of the feed water pump 23, the cooling performance of the reactor core 2 is improved. However, this case also has the following technical problems.

原子炉の事故時における原子炉容器１内の圧力及び温度
の変化を第３図に示す、事故時には配管破断部からの一
次冷却材４の流出及び原子炉停止にともなう炉心２での
発熱（崩壊熱）の低下によって圧力Ｐ、が低下する。こ
の圧力低下にともなって飽和温度ＴＶ、も低下し、一次
冷却材４の温度Ｔｏが飽和温度Ｔ□と等しくなり沸騰を
開始する。Figure 3 shows the changes in pressure and temperature inside the reactor vessel 1 during a nuclear reactor accident. At the time of an accident, primary coolant 4 flows out from the pipe rupture and heat generation (collapse) occurs in the reactor core 2 due to reactor shutdown. The pressure P decreases due to the decrease in heat). As the pressure decreases, the saturation temperature TV also decreases, and the temperature To of the primary coolant 4 becomes equal to the saturation temperature T□, and boiling begins.

一次冷却材４の沸騰開始後にはその温度Ｔ。は飽和温度
Ｔ□に従って低下する。なお、一次冷却材４の温度Ｔ。After the primary coolant 4 starts boiling, its temperature T. decreases according to the saturation temperature T□. In addition, the temperature T of the primary coolant 4.

は冷却水貯水槽２１内の圧力に対する飽和温度Ｔ□より
はるかに高く、冷却水２２の温度をＴＷ、まで加熱して
も、前述の如く熱衝撃の発生を回避できない、今、一次
冷却材４の比熱をＣ１゜、流量をＷｃとし、冷却水２２
の温度をＴＷ。is much higher than the saturation temperature T□ for the pressure in the cooling water storage tank 21, and even if the temperature of the cooling water 22 is heated to TW, the occurrence of thermal shock cannot be avoided as described above. The specific heat of the cooling water is C1°, the flow rate is Wc, and the cooling water 22
The temperature of TW.

比熱をＣ工、流量をＷＷとすると、混合熱の温度Ｔ、は
熱量保存則から次式のようになる。When the specific heat is C and the flow rate is WW, the temperature T of the mixed heat is expressed by the following equation based on the law of conservation of heat.

Ｃ，。Ｔ　（！　Ｗ（１＋　Ｃｒｗ　Ｔ　ｌ−Ｗ　Ｗ　
＝ｃ　Ｐ−Ｔ　−（ｗ０＋　ｗｗ　）・・・・・・（１
） 各比熱Ｃ，，，Ｃ，□及びＣ０はほぼ等しくゝから（１
）式は次のように近似できる。C. T (! W(1+ Crw T l-W W
=c P−T −(w0+ww)・・・・・・(1
) Each specific heat C, , C, □ and C0 are approximately equal from も to (1
) can be approximated as follows.

Ｔ、＝　（ＴｃＷｏ＋Ｔ、ＷＷ）／　（ｗｃ＋ｗｗ）　
・（２）一般に、ポンプ特性は第４図（Ａ）に示すよう
に吐出圧力が低下すると吐出流量が増加する。原子炉容
器１内の圧力Ｐ、は第３図に示したように事故発生後低
下し、したがって、給水ポンプ２３の吐出圧力も圧力Ｐ
ｖの低下にそって低下するため、冷却水２２の流量Ｗｗ
は第４図（Ｂ）に示すように増加する。一方、一次冷却
材４は原子炉容器１から出て原子炉容器１にもどるため
、その流量Ｗｃは圧力Ｐｖに関係なく第４図（Ｂ）に示
すようにほぼ一定となる。さらに、冷却水２２の加熱源
である一次冷却材４の温度Ｔ０は第３図及び第４図（Ｃ
）に示すように低下する。すなわち、（２）式からも明
らかなように、熱源の温度Ｔ。T, = (TcWo+T, WW)/(wc+ww)
- (2) In general, as shown in FIG. 4(A), the pump characteristic is that when the discharge pressure decreases, the discharge flow rate increases. As shown in FIG. 3, the pressure P inside the reactor vessel 1 decreases after the accident occurs, and therefore the discharge pressure of the feed water pump 23 also decreases
The flow rate Ww of the cooling water 22 decreases as v decreases.
increases as shown in FIG. 4(B). On the other hand, since the primary coolant 4 leaves the reactor vessel 1 and returns to the reactor vessel 1, its flow rate Wc becomes approximately constant as shown in FIG. 4(B) regardless of the pressure Pv. Furthermore, the temperature T0 of the primary coolant 4, which is the heating source of the cooling water 22, is shown in FIGS. 3 and 4 (C
). That is, as is clear from equation (2), the temperature T of the heat source.

が低下し、低温水の流量ｗｗが増加すると、混合後の温
度Ｔ、は第４図（Ｃ）に示すように急激に低下する。し
たがって、混合後の温度Ｔ、を高く保持するためには、
混合後の温度Ｔ、を検出し。decreases and the flow rate ww of low-temperature water increases, the temperature T after mixing rapidly decreases as shown in FIG. 4(C). Therefore, in order to maintain a high temperature T after mixing,
Detect the temperature T after mixing.

一次冷却材４の流量Ｗ０及び冷却水２２の流量Ｗｗを調
節しなければならないにのような流量制御は流量調節弁
によって実施できる。Flow rate control such as the need to adjust the flow rate W0 of the primary coolant 4 and the flow rate Ww of the cooling water 22 can be performed by a flow rate control valve.

以上の検討結果から１本発明では、 （１）一次冷却材と冷却水とを冷却水給水ポンプの下流
部で混合し、かつ、混合後の温度を検出して一次冷却材
及び冷却水の流量を調節することにより、混合後の温度
を冷却水貯水槽内の圧力に対する飽和温度より高く、一
次冷却材の温度以下である任意の温度にできるようにす
る。From the above study results, the present invention (1) mixes the primary coolant and cooling water downstream of the cooling water supply pump, detects the temperature after mixing, and detects the flow rate of the primary coolant and cooling water. By adjusting the temperature, the temperature after mixing can be set to an arbitrary temperature that is higher than the saturation temperature for the pressure in the cooling water storage tank and lower than the temperature of the primary coolant.

上記方法において、冷却水２２の流量Ｗｗを減らすと原
子炉容器１に供給する正味の冷却水量が減少することに
なる（なぜならば、一次冷却材４は原子炉容器１と高温
配管ルートＤを再循環するのみであるから）。したがっ
て、本発明では、炉心２の冷却をより効果的に実施する
ために、（２）第１に一次冷却材の流量を調節し、一次
冷却材の流量調節が限界に達した後、冷却水の流量を調
節する。In the above method, when the flow rate Ww of the cooling water 22 is reduced, the net amount of cooling water supplied to the reactor vessel 1 is reduced (because the primary coolant 4 is recirculated between the reactor vessel 1 and the high-temperature piping route D). (because it only circulates). Therefore, in the present invention, in order to cool the core 2 more effectively, (2) first, the flow rate of the primary coolant is adjusted, and after the flow rate adjustment of the primary coolant reaches its limit, the cooling water is Adjust the flow rate.

上記の方法において必要以上に冷却水２２の温度を上昇
させることは、炉心２の冷却性能を減少させることを意
味する。したがって、本発明では、（３）原子炉容器内
の圧力が原子炉容器を損傷しない圧力まで低下した後に
は、一次冷却水の混合を停止し、この圧力以上の場合に
のみ混合後の温度を原子炉容器材料の低温損傷下限温度
以上に制御する。すなわち、原子炉容器内の圧力が２０
気圧以上では混合後の温度を２００℃以上に制御し、２
０気圧以下では一次冷却水の流量を零とする。In the above method, increasing the temperature of the cooling water 22 more than necessary means reducing the cooling performance of the core 2. Therefore, in the present invention, (3) mixing of the primary cooling water is stopped after the pressure inside the reactor vessel has decreased to a pressure that does not damage the reactor vessel, and the temperature after mixing is reduced only when the pressure is above this pressure. Control the temperature above the lower limit for low-temperature damage to reactor vessel materials. In other words, the pressure inside the reactor vessel is 20
At atmospheric pressure or higher, the temperature after mixing should be controlled at 200°C or higher, and 2
Below 0 atmospheric pressure, the flow rate of the primary cooling water is set to zero.

以上、述べた本発明における混合後の温度Ｔ。The temperature T after mixing in the present invention described above.

の具体的な制御方法を第５図を用いて説明する。A specific control method will be explained using FIG.

原子炉容器１内の圧力Ｐ、は第５図（Ａ）に示すように
事故発生後低下し、この圧力低下にともなって一次冷却
材４の温度Ｔ０も第５図（Ｂ）に示すように低下する。The pressure P in the reactor vessel 1 decreases after the accident, as shown in FIG. 5(A), and with this pressure decrease, the temperature T0 of the primary coolant 4 also decreases as shown in FIG. 5(B). descend.

冷却水２２の注水開始後、冷却水２２の流量ＷＷが第５
図（Ｃ）に示すように圧力ＰＶの低下に従って増加し、
一次冷却材４の温度Ｔ。は第５図（Ｂ）に示すように低
下するため、一次冷却材５の流量Ｗｃを第５図（Ｃ）に
示すように増加させて、冷却水２２と一次冷却材４との
混合後の温度Ｔ、を原子炉圧力容器材料の低温損傷下限
温度である２００℃以上に保持する。After the start of water injection of the cooling water 22, the flow rate WW of the cooling water 22 reaches the fifth level.
As shown in figure (C), it increases as the pressure PV decreases,
Temperature T of the primary coolant 4. decreases as shown in FIG. 5(B), so the flow rate Wc of the primary coolant 5 is increased as shown in FIG. 5(C), and the The temperature T is maintained at 200° C. or higher, which is the lower limit temperature for low-temperature damage to the reactor pressure vessel material.

一次冷却材４の流量Ｗ。が上限値Ｗ　Ｑ　ｌ　ｍ　１１
１１に達した後には、低温の冷却水２２の流量ＷＷを減
少させ、混合後の温度Ｔ、を２００℃以下に保持する。Flow rate W of primary coolant 4. is the upper limit W Q l m 11
11, the flow rate WW of the low-temperature cooling water 22 is reduced to maintain the temperature T after mixing at 200° C. or lower.

原子炉容器１内の圧力ＰＹが原子炉容器の損傷しない圧
力２０気圧以下になると冷却水２２の加熱が不要となる
ため、一次冷却材４の流量Ｗ。を零とし、冷却水２２の
流量ＷＷを最大流量Ｗ□１．。When the pressure PY in the reactor vessel 1 becomes 20 atmospheres or less at which the reactor vessel is not damaged, heating of the cooling water 22 becomes unnecessary, so the flow rate W of the primary coolant 4 is reduced. is set to zero, and the flow rate WW of the cooling water 22 is set to the maximum flow rate W□1. .

にして炉心２の冷却能力を向上させる。to improve the cooling capacity of the core 2.

〔発明の実施例〕[Embodiments of the invention]

次に、本発明の一実施例を第１図により説明する。本実
施例の特徴は、冷却水２２と一次冷却材４との混合容器
２６にエゼクタを用いたことである。事故の発生を検出
して給水ポンプ２３を駆動するとともに流量調節弁２９
と流量調節弁２５を全開する。冷却水２２は給水ポンプ
２３で逆止弁２４を通り、ノズル２７から混合容器２６
内に高速噴出される。この冷却水２２の高速噴流によっ
て、沸騰水型原子炉におけるジェットポンプと同じ原理
で、高温の一次冷却材４が高温配管２８を通して吸入さ
れ混合容器２６内で低温の冷却水２２と混合し、原子炉
容器１に供給される。この時、混合後の温度Ｔ、を温度
検出器３２で検出し、温度Ｔ３が設定値（以下の実施例
では設定値を２００℃としているが、２００℃以上あれ
ば良い）となるように、制御袋！！３１を介して流量調
節弁２９で一次冷却材４の流量を制御する。第５図で述
べたように、流量調節弁２９の制御が限界値に達した後
には、流量調節弁２９と同様の方法で流量調節弁２５を
制御する。さらに、第５図で述べたように、原子炉容器
１の圧力検出器３３で検出された気力が設定値以下にな
ると、制御装置３１を介して流量調節弁２９を全閉して
一次冷却材４の流量を零とし、流量調節弁２５を全開さ
せ炉心２の冷却能力を向上させる。本実施例では、一次
冷却水４をホットレグ５から引き出し混合容器２６で冷
却水２２と混合した後、コールドレグ９に注入している
が、直接に原子炉容器１から引き出すことも、加圧器１
０もしくはコールドレグ９から引き出すことも可能であ
り、また、直接圧力容器１にもしくはホットレグ５に注
入することも可能である。Next, one embodiment of the present invention will be described with reference to FIG. A feature of this embodiment is that an ejector is used as the mixing container 26 for the cooling water 22 and the primary coolant 4. Detects the occurrence of an accident and drives the water supply pump 23, as well as the flow control valve 29.
and fully open the flow control valve 25. Cooling water 22 passes through a check valve 24 with a water supply pump 23, and flows from a nozzle 27 to a mixing container 26.
It is ejected inside at high speed. By this high-speed jet of cooling water 22, the high-temperature primary coolant 4 is sucked in through the high-temperature piping 28 and mixed with the low-temperature cooling water 22 in the mixing vessel 26, based on the same principle as a jet pump in a boiling water reactor. It is supplied to the furnace vessel 1. At this time, the temperature T3 after mixing is detected by the temperature detector 32, and the temperature T3 is set at a set value (in the examples below, the set value is 200°C, but it is sufficient if it is 200°C or higher). Control bag! ! The flow rate of the primary coolant 4 is controlled by the flow rate control valve 29 via the primary coolant 31 . As described in FIG. 5, after the control of the flow control valve 29 reaches the limit value, the flow control valve 25 is controlled in the same manner as the flow control valve 29. Furthermore, as described in FIG. 5, when the pressure detected by the pressure detector 33 of the reactor vessel 1 becomes less than the set value, the flow rate control valve 29 is fully closed via the control device 31 and the primary coolant is 4 is set to zero, and the flow rate control valve 25 is fully opened to improve the cooling capacity of the core 2. In this embodiment, the primary cooling water 4 is drawn out from the hot leg 5, mixed with the cooling water 22 in the mixing vessel 26, and then injected into the cold leg 9. However, it is also possible to draw it out directly from the reactor vessel 1, or
It is also possible to withdraw from the 0 or cold leg 9 or to inject directly into the pressure vessel 1 or into the hot leg 5.

本実施例の重要な構成要素である混合容器２６として用
いるエゼクタの構造及び特性を第６図及び第７図に示す
。エゼクタはノズル２７から高速で流体を噴出して周囲
の流体を吸入するものである。この時、駆動水である冷
却水２２の流量ＷＷと吸入される一次冷却材４の流量Ｗ
ｃとの比（Ｗ、／Ｗ、）は、従来技術によるエゼクタの
ように、ノズル２７の流路面積Ａ、とスロート３０の流
路面積Ａｔｋとの比（Ａｔｈ／Ａ−）が一定の場合には
、第７図（Ａ）　ｃこ示すように、流量ＷＷによらずほ
ぼ一定となる。この場合、第５図（Ｃ）において冷却水
２２の流量Ｗｗを配管に設けた流量調節弁で減少させる
と一次冷却水４の流量Ｗ。The structure and characteristics of the ejector used as the mixing container 26, which is an important component of this embodiment, are shown in FIGS. 6 and 7. The ejector ejects fluid from the nozzle 27 at high speed and sucks in surrounding fluid. At this time, the flow rate WW of the cooling water 22 which is the driving water and the flow rate W of the primary coolant 4 to be sucked.
The ratio (W, /W,) to c is when the ratio (Ath/A-) between the flow path area A of the nozzle 27 and the flow path area Atk of the throat 30 is constant, as in the ejector according to the prior art. As shown in FIG. 7(A), the flow rate remains almost constant regardless of the flow rate WW. In this case, in FIG. 5(C), when the flow rate Ww of the cooling water 22 is decreased by the flow rate control valve provided in the pipe, the flow rate W of the primary cooling water 4 is reduced.

も減少してしまい、初期の目的を達成できない。will also decrease, making it impossible to achieve the initial objective.

そこで、本実施例におけるエゼクタでは、流量調節弁２
５によって、ノズル２７の流路面積Ａ、を可変とする。Therefore, in the ejector in this embodiment, the flow control valve 2
5, the flow path area A of the nozzle 27 is made variable.

流路面積Ａ、を小さくして流路面積比（Ａ、、／Ａ、）
　　を増加させると第７図（Ｂ）に示すように流量比（
Ｗ、／Ｗ、）が増加する。すなわち、流量調節弁２５を
絞ってノズル２７の流路面積Ａ、を小さくすると、冷却
水２２側の流動抵抗が増加し流量ＷＷが減少するが、流
路面積比（Ａ、に／Ａ、）が増加するため一次冷却材４
の流量Ｗ。は減少しない。Reduce the flow path area A to obtain the flow path area ratio (A, , /A,)
As shown in Figure 7 (B), when increasing the flow rate ratio (
W,/W,) increases. That is, when the flow rate control valve 25 is throttled to reduce the flow area A of the nozzle 27, the flow resistance on the cooling water 22 side increases and the flow rate WW decreases, but the flow area ratio (A, to/A,) The primary coolant 4
The flow rate W. does not decrease.

本実施例における制御特性を第８図に示す、事故発生の
検出時間遅れ及び駆動ポンプ２３の起動時間遅れにより
事故発生後ｔ□で流量調節弁２５及び２９が全開され、
冷却水２２と一次冷却材４が混合された後、原子炉容器
１への供給が開始される。この時には、流量調節弁２９
が全開しているので、混合後の温度Ｔ、は設定値より若
干高くなっているが、温度検出器３２の検出値に基づき
制御装置３１により流量調節弁２９を絞り一次冷却材４
の流量Ｗｃを減少させ混合後の温度Ｔ、を設定値に保持
する。原子炉容器１内の圧力Ｐ、の低下にともなって、
第４図（Ａ）に示したポンプ特性から冷却水２２の流量
Ｗｗが増加し、第６図（Ａ）に示したエゼクタの特性か
ら一次冷却材４の流量Ｗ。も増加するが、混合後の温度
Ｔ２を設定値に保つように制御装置３１を介して流量調
節弁２９により一次冷却材４の流量Ｗ０を制御する。The control characteristics in this embodiment are shown in FIG. 8. Due to the delay in the detection time of the occurrence of an accident and the delay in the start-up time of the drive pump 23, the flow rate control valves 25 and 29 are fully opened at t□ after the occurrence of the accident.
After the cooling water 22 and the primary coolant 4 are mixed, supply to the reactor vessel 1 is started. At this time, the flow rate control valve 29
is fully open, the temperature T after mixing is slightly higher than the set value, but the control device 31 throttles the flow rate control valve 29 based on the detected value of the temperature detector 32.
The temperature T after mixing is maintained at the set value by decreasing the flow rate Wc. As the pressure P inside the reactor vessel 1 decreases,
The flow rate Ww of the cooling water 22 increases based on the pump characteristics shown in FIG. 4(A), and the flow rate W of the primary coolant 4 increases based on the ejector characteristics shown in FIG. 6(A). However, the flow rate W0 of the primary coolant 4 is controlled by the flow control valve 29 via the control device 31 so as to maintain the temperature T2 after mixing at the set value.

時間ｔ２で流量調節弁２９の弁開度が１００％となり一
次冷却材４の流量Ｗ０の制御が限界となった後、制御装
置３１を介して流量調節弁２５を絞リーノズル２７の流
路面積Ａ、を小さくして冷却水２２の流量ＷＷを減少さ
せる。この時、第７図（Ｂ）に示したように流路面積比
（Ａ、、／Ａ、）が増加するため、駆動水である冷却水
２２の流量ＷＷが減少しても一次冷却材４の流量Ｗｔ３
は減少しない、さらに、原子炉容器１内の圧力Ｐｖが設
定値以下になると流量調節弁２９を全閉し、流量調節弁
２５を全開する。At time t2, the valve opening degree of the flow rate control valve 29 becomes 100%, and after the control of the flow rate W0 of the primary coolant 4 reaches its limit, the flow rate control valve 25 is throttled through the control device 31 to reach the flow path area A of the Lee nozzle 27. , to reduce the flow rate WW of the cooling water 22. At this time, as shown in FIG. 7(B), the flow path area ratio (A, , /A,) increases, so even if the flow rate WW of the cooling water 22, which is the driving water, decreases, the primary coolant 4 Flow rate Wt3
does not decrease.Furthermore, when the pressure Pv inside the reactor vessel 1 becomes less than the set value, the flow rate control valve 29 is fully closed and the flow rate control valve 25 is fully opened.

第９図に本発明の他の実施例を示す０本実施例の特徴は
一次冷却材４を高温ポンプ３５で混合容器２６に供給す
ることである。冷却水２２と一次冷却材４との混合後の
温度Ｔ、の制御方法は第１図に示した実施例の場合と同
様である１本実施例では、−水冷却材４を高温ポンプ３
５で駆動するため、高温ポンプ３５の入口における一次
冷却水の温度はポンプ内での蒸気発生によるポンプの損
傷を防止するために飽和温度未満にしなければならない
、このため、本実施例ではバイパス配管３６を通して給
水ポンプ２３で供給される低温の冷却水２２の一部を温
水ポンプ３５の上流部に供給する。圧力検出器４３で温
水ポンプ３５の入口の圧力Ｐ、を測定し、この圧力Ｐ２
に対する飽和温度Ｐ４を計算し、温度検出器４２の検出
温度Ｔ２が飽和温度Ｔ４より低くなるように温度制御装
置４１を介して流量調節弁３７を制御する。飽和温度Ｔ
、、と温度Ｔ２との差（Ｔ　Ｐ　ａ　−Ｔ　Ｐ　）は５
〜１０℃にすればよい０本実施例における制御特性は第
８図に示した第１図の実施例の場合と同様であるが、−
水冷却材４の流量Ｗ０は第５図（Ｃ）に示すようになる
。Another embodiment of the present invention is shown in FIG. 9. The feature of this embodiment is that the primary coolant 4 is supplied to the mixing vessel 26 by a high temperature pump 35. The method of controlling the temperature T after mixing the cooling water 22 and the primary coolant 4 is the same as in the embodiment shown in FIG.
5, the temperature of the primary cooling water at the inlet of the high-temperature pump 35 must be below the saturation temperature to prevent damage to the pump due to steam generation within the pump.For this reason, in this embodiment, the bypass piping is A part of the low-temperature cooling water 22 supplied by the water supply pump 23 is supplied to the upstream portion of the hot water pump 35 through 36 . The pressure detector 43 measures the pressure P at the inlet of the hot water pump 35, and this pressure P2
A saturation temperature P4 is calculated for the temperature, and the flow rate regulating valve 37 is controlled via the temperature control device 41 so that the temperature T2 detected by the temperature detector 42 is lower than the saturation temperature T4. Saturation temperature T
, , and the temperature T2 (T P a - T P ) is 5
The control characteristics in this embodiment are the same as those in the embodiment shown in FIG. 1 shown in FIG. 8, but -
The flow rate W0 of the water coolant 4 is as shown in FIG. 5(C).

第１０図に本発明のさらに他の実施例を示す。FIG. 10 shows still another embodiment of the present invention.

本実施例の特徴は、高温ポンプ３５を冷却水２２と一次
冷却材４とを混合する混合容器２６の下流部に設けたこ
とである。混合後の温度Ｔ、の制御方法は第１図に示し
た実施例の場合と同様であるが、第９図に示した実施例
で説明したように、高温ポンプ３５を保護するために入
口の温度Ｔ、を圧力Ｐ、に対する飽和温度１．１未満に
制御しなければならない。また、−水冷却材４の混合容
器２６への供給を可能とするために、混合容器２６の圧
力Ｐ、を原子炉容器１内の圧力Ｐｙより低くしなければ
ならない、このため１本実施例では、給水ポンプ２３に
バイパス配管３６と流量調節弁３７を設け、圧力制御袋
Ｗ１４４で圧力Ｐ、で圧力Ｐｖ未満となるように給水ポ
ンプ２３の吐出圧力を制御する。The feature of this embodiment is that the high temperature pump 35 is provided downstream of the mixing container 26 that mixes the cooling water 22 and the primary coolant 4. The method of controlling the temperature T after mixing is the same as in the embodiment shown in FIG. 1, but as explained in the embodiment shown in FIG. The temperature T, must be controlled below the saturation temperature 1.1 for the pressure P. Furthermore, in order to enable supply of the water coolant 4 to the mixing vessel 26, the pressure P in the mixing vessel 26 must be lower than the pressure Py in the reactor vessel 1. For this reason, one embodiment Now, the water supply pump 23 is provided with a bypass pipe 36 and a flow rate control valve 37, and the discharge pressure of the water supply pump 23 is controlled so that the pressure P in the pressure control bag W144 is less than the pressure Pv.

本実施例の制御特性を第１１図に示す。給水ポンプ２３
の吐出圧力を低くし混合容器２６の圧力Ｐ、を原子炉容
器１の圧力Ｐｖより低くする。圧力Ｐ、が高い場合には
、圧力制御装置４４で流量調節弁３７を開き、第４図（
Ａ）に示したポンプ特性から明らかなようにポンプ吐出
流量を増加させ吐出圧力を低下させる。逆に、圧力Ｐ、
が低過ぎる場合には流量調節弁３７を絞ればよい、この
ようにして、圧力Ｐ、を圧力Ｐ、より低くすると、圧力
差（ｐ、−ｐ、）によって原子炉容器１から高温の一次
冷却材４が混合容器２６に供給される。FIG. 11 shows the control characteristics of this embodiment. Water supply pump 23
The discharge pressure of the reactor vessel 1 is lowered to make the pressure P of the mixing vessel 26 lower than the pressure Pv of the reactor vessel 1. When the pressure P is high, the flow rate control valve 37 is opened by the pressure control device 44, and as shown in FIG.
As is clear from the pump characteristics shown in A), the pump discharge flow rate is increased and the discharge pressure is decreased. On the contrary, the pressure P,
If P is too low, the flow rate control valve 37 can be throttled down. In this way, by lowering the pressure P, the pressure difference (p, -p,) causes high temperature primary cooling from the reactor vessel 1. Material 4 is fed into mixing vessel 26 .

この−水冷却材４は給水ポンプ２３で混合容器２６に供
給される冷却水２２と混合した後、温水ポンプ３５で原
子炉容器１に供給される。この時、温水ポンプ３５を保
護するために、混合後の温度Ｔ、は圧力Ｐ、に対する飽
和温度Ｔ、、よりも少なくとも５〜１０℃低くなるよう
に制御される。この温度差（Ｔ−、−’ｒ、）が、万一
、５〜１０℃より小さくなった場合には、制御装置３１
で流量調節弁２９を絞り一次冷却材４の流量を減少させ
るかもしくは流量調節弁２５を開き冷却水２２の流量を
増加させればよい。この制御方法は他の実施例の場合と
同様である。This -water coolant 4 is mixed with the cooling water 22 supplied to the mixing vessel 26 by the water supply pump 23, and then is supplied to the reactor vessel 1 by the hot water pump 35. At this time, in order to protect the hot water pump 35, the temperature T after mixing is controlled to be at least 5 to 10 degrees Celsius lower than the saturation temperature T with respect to the pressure P. If this temperature difference (T-, -'r,) becomes smaller than 5 to 10°C, the control device 31
The flow rate adjustment valve 29 may be throttled to reduce the flow rate of the primary coolant 4, or the flow rate adjustment valve 25 may be opened to increase the flow rate of the cooling water 22. This control method is the same as in the other embodiments.

次に、上述の流量制御を行なう制御装置３１の詳細につ
いて説明する。上述各実施例の基本となる制御装置の構
成の一例を第１２図に示す。原子炉容器１内の圧力検出
器３３の信号は、演算回路４５により圧力値に変換され
る。この信号と、設定回路４６からの信号（この場合、
制御をするかしないかのしきい値である２０気圧という
圧力値）が比較回路４７に入り、Ｐ、＞２０気圧であれ
ば論理上の偽信号、Ｐｖ＜２０気圧であれば論理上の真
信号が、信号４８として発せられる。一方。Next, details of the control device 31 that performs the above-described flow rate control will be explained. FIG. 12 shows an example of the configuration of a control device that is the basis of each of the above embodiments. The signal from the pressure detector 33 inside the reactor vessel 1 is converted into a pressure value by the arithmetic circuit 45. This signal and the signal from the setting circuit 46 (in this case,
The pressure value of 20 atm, which is the threshold value for whether or not to perform control, enters the comparator circuit 47, and if P > 20 atm, it is a logical false signal, and if Pv < 20 atm, it is a logical true signal. A signal is emitted as signal 48. on the other hand.

混合後の温度を検出する温度検出器３２の信号は、演算
回路４９に入り温度値に変換される。この信号は分岐さ
れ、それぞれが設定回路５０からの信号（この場合は、
設定温度２００℃）とともに比較回路５１．５２に入る
。比較回路５１では、’ｌ＞２００の場合に論理上の真
信号が、それ以外の場合は偽信号が信号５３として発信
される。A signal from the temperature detector 32 that detects the temperature after mixing enters an arithmetic circuit 49 and is converted into a temperature value. This signal is branched, and each signal is a signal from the setting circuit 50 (in this case,
The set temperature is 200° C.) and enters the comparator circuits 51 and 52. The comparison circuit 51 outputs a logically true signal as a signal 53 when 'l>200, and a false signal in other cases.

また比較回路５２では、Ｔ、＜２００の場合に論理上の
真信号が、それ以外の場合は偽信号が信号５４として発
信される。信号５３は、分岐され、その一方は、流量調
整弁２５の開示上限を論理上の真値で示す信号（例えば
、上限リミットスイッチ信号などで、上限の場合のみ真
信号を発信する）５５が反転回路５６で反転された結果
と共にＡＮＤ回路５７に入り論理積がとられ、真偽値信
号５８が発信される１分岐された他方の信号は、弁２５
の開度上限を論理上の真値で示す信号５５と共にＡＮＤ
回路５９に入り、この回路での論理積の結果として真偽
値信号６０が発信される。これらの真偽値信号５８．６
０は、それぞれ、前述の容器圧力が２０気圧以下である
かどうかの比較結果信号４８が反転回路６１で反転され
た結果信号６２と共に、ＡＮＤ回路６３．６４に入る。Further, the comparison circuit 52 outputs a logically true signal as the signal 54 when T<200, and a false signal in other cases. The signal 53 is branched, and one of them is a signal indicating the upper limit of the disclosure of the flow rate regulating valve 25 as a logical true value (for example, an upper limit switch signal, which transmits a true signal only when the upper limit is reached), and the signal 55 is inverted. The other signal, which is branched into one, enters an AND circuit 57 together with the inverted result in the circuit 56 and performs a logical product to generate a truth value signal 58.
AND with signal 55 indicating the upper limit of the opening degree as a logical true value.
The circuit 59 is entered and a truth value signal 60 is generated as a result of the AND operation in this circuit. These truth value signals 58.6
0 enters the AND circuits 63 and 64 together with the result signal 62 obtained by inverting the comparison result signal 48 as to whether the container pressure is 20 atmospheres or less in the inverting circuit 61, respectively.

ＡＮＤ回路６３．６４の論理積の結果は、それぞれ、制
御弁２５の開動作を行なわせるガバナ６５と制御弁２９
の閉動作を行なわせるガバナ６６に伝えられる。The results of the AND circuits 63 and 64 are the governor 65 and the control valve 29 that open the control valve 25, respectively.
The signal is transmitted to the governor 66, which causes the closing operation to be performed.

この回路により、Ｔ、＞２００℃でＰ、ン２２０気圧の
場合（すなわち、温度制御が必要な圧力であり混合後の
温度が設定値より高い場合）、弁２５の開度上限でなけ
れば、弁２５に開動作（低温水の増加動作）が指示され
、弁２５の開度上限であれば、弁２９に閉動作（すなわ
ち、高温水側の流量減少動作）が指示され、混合後の温
度が設定値に制御される。一方、”ｌ＜２００℃の場合
に真値信号が発信される信号５４は、分岐され、一方は
、流量制御弁２９の開度上限を論理上の真値で示す信号
６７が反転回路６８で反転された結果と共にＡＮＤ回路
６９に入り、他方は、前記信号６７と共にＡＮＤ回路７
（Ｈこ入る。ＡＮＤ回路６９．７０における論理積の結
果は、それぞれ、前述の容器圧力が２０気圧以下である
かどうかの比較結果信号４８が反転された結果６２と共
に、ＡＮＤ回路７３．７４に入る。ＡＮＤ回路７３゜７
４の論理積の結果は、それぞれ、制御弁２９の開動作を
行なわせるガバナ７５と制御弁２５の閉動作を行なわせ
るガバナ７６に伝えられる。この回路の動作により、Ｔ
、＜２００℃でＰ、＞２０気圧の場合（すなわち、温度
制御が必要な圧力であり、混合後の温度が低い場合）、
弁２９の開度上限でなければ、弁２９に開動作（高温水
の流量増加動作）が指示され、弁２９が開度上限であれ
ば、弁２５に閉動作（低温水の流量減少動作）が指示さ
れる。この結果、混合後の温度が設定値に制御される６
また。ｐ、＜２０気圧であれば、比較回路４７から発信
される真値信号４８が直接、弁２９の閉動作を指示する
ガバナ７７と弁２５の開動作を指示するガバナ７８に伝
えられているので、温度制御は解除される。With this circuit, if T is >200°C and P is 220 atm (that is, the pressure requires temperature control and the temperature after mixing is higher than the set value), if the opening degree of the valve 25 is not at the upper limit, The valve 25 is instructed to open (increase the low temperature water), and if the opening degree of the valve 25 is at the upper limit, the valve 29 is instructed to close (in other words, the flow rate decreases on the high temperature water side), and the temperature after mixing increases. is controlled to the set value. On the other hand, the signal 54 from which the true value signal is transmitted when l<200°C is branched, and the signal 67 indicating the upper limit of the opening degree of the flow rate control valve 29 as a logical true value is transmitted to the inverting circuit 68. The inverted result is input to an AND circuit 69, and the other signal is input to an AND circuit 7 together with the signal 67.
(H enters. The results of the logical product in the AND circuits 69 and 70 are respectively input to the AND circuits 73 and 74 together with the result 62 obtained by inverting the comparison result signal 48 as to whether the container pressure is 20 atmospheres or less. Enter.AND circuit 73°7
The results of the logical product of 4 are transmitted to the governor 75, which causes the control valve 29 to open, and the governor 76, which causes the control valve 25 to close. Due to the operation of this circuit, T
, P at <200°C, >20 atm (i.e., at a pressure where temperature control is required and the temperature after mixing is low),
If the opening degree of the valve 29 is not at the upper limit, the valve 29 is instructed to open (operation to increase the flow rate of high-temperature water), and if the opening degree of the valve 29 is at the upper limit, the valve 25 is instructed to close the operation (operation to decrease the flow rate of low-temperature water). is instructed. As a result, the temperature after mixing is controlled to the set value6.
Also. If p is <20 atm, the true value signal 48 sent from the comparator circuit 47 is directly transmitted to the governor 77, which instructs the closing operation of the valve 29, and the governor 78, which instructs the opening operation of the valve 25. , temperature control is canceled.

以上説明した回路及び動作により、前述の温度制御が達
成される。The above-described temperature control is achieved by the circuit and operation described above.

また、第１０図の実施例において、混合後の温度を、設
定値と混合後の圧力Ｐ、における飽和温度より５〜１０
℃低い温度のいずれか低い温度に制御する場合の制御回
路は、第１３図に示す構成となる。この回路においては
、混合後の圧力を計測した信号４３は、演算回路７９に
入り圧力値に変換される。この値は、演算回路８０に入
り、圧力の飽和温度−（５〜１０℃）の温度値になる。In addition, in the example shown in FIG.
A control circuit for controlling the temperature to the lower temperature of 0.degree. C. or lower has a configuration shown in FIG. 13. In this circuit, a signal 43 measuring the pressure after mixing enters an arithmetic circuit 79 and is converted into a pressure value. This value is input to the arithmetic circuit 80 and becomes a temperature value of the pressure saturation temperature - (5 to 10°C).

この温度値信号は、温度設定回路５０からの設定値（こ
の場合は２００℃と共に低位設定回路８１に入る。）低
位設定回路８１では両者の比較後、低い値の方を設定値
として発信し、比較回路５１゜５２に伝える。この部分
以外の回路構成は第１２図に示す回路と同一であり、こ
の回路の作動により、混合後の温度は、設定値か混合位
置での圧力Ｐ、における飽和温度より５〜１０℃低い温
度のどちらか低い温度に制御される。This temperature value signal is input to the low-level setting circuit 81 together with the set value from the temperature setting circuit 50 (in this case, 200° C.).The low-level setting circuit 81 compares the two, and then transmits the lower value as the set value. The information is transmitted to comparison circuits 51 and 52. The circuit configuration other than this part is the same as the circuit shown in Fig. 12, and by the operation of this circuit, the temperature after mixing is 5 to 10 degrees Celsius lower than the saturation temperature at the set value or the pressure P at the mixing position. The temperature is controlled to the lower of the two.

〔発明の効果〕〔Effect of the invention〕

本発明によれば、高温の一次冷却材と低温の冷却水とを
直接混合し、冷却水の温度を冷却水貯水槽の飽和温度よ
り高い任意の温度に保持できるので、冷却水の注入によ
る熱衝撃の発生を防止できる効果がある。本発明では、
冷却水の加熱に高温の一次冷却材を用いるから新たな加
熱源が不要であり、エゼクタもしくはポンプ及び制御装
置の追加程度で実現でき、経済的である。According to the present invention, the temperature of the cooling water can be maintained at an arbitrary temperature higher than the saturation temperature of the cooling water storage tank by directly mixing the high-temperature primary coolant and the low-temperature cooling water. This has the effect of preventing the occurrence of impact. In the present invention,
Since a high-temperature primary coolant is used to heat the cooling water, a new heating source is not required, and it can be realized by adding an ejector or pump and a control device, which is economical.

また、本発明によれば、一次冷却材を再循環させるため
、炉心の冷却効果を向上させることが可能である。Further, according to the present invention, since the primary coolant is recirculated, it is possible to improve the cooling effect of the core.

【図面の簡単な説明】[Brief explanation of the drawing]

第１図は本発明による一実施例を示すブロック図、第２
図は従来技術を説明するブロック図、第３図は事故時の
圧力及び温度変化の説明図、第４図は一般的なポンプ特
性と流量及び温度変化の説明図、第５図は本発明による
制御特性の説明図。 第６図は第１図の部分詳細図、第７図は第６図の特性を
示す図、第８図は第１図の制御特性を示す図、第９図は
本発明による他の実施例の構成を示すブロック図、第１
０図は本発明によるさらに他の実施例の構成を示すブロ
ック図、第１１図は第１０図の制御方法を説明する図、
第１２＠は混合後の温度制御を達成する基本的な回路構
成を示す図、第１３図は第１０図しこ示された実施例に
おける温度 制御を達成する回路構成を示す図である。 １・・・原子炉容器、２・・・炉心、３・・・ダウンカ
マ、４・・・一次冷却材、５・・・ボットレグ、６・・
・蒸気発生器。 ７・・・伝熱管、８・・・循環ポンプ、９・・・コール
ドレグ、１０・・・加圧器、１５・・・二次冷却材、２
１・・・冷却水貯水槽、２２・・・冷却水、２３・・・
給水ポンプ、２４・・・逆止弁、２５・・・流量調節弁
、２６・・・混合容器。 ２７・・・エゼクタノズル、２８・・・高温配管、２９
・・・流量調節弁、３０・・・スロート、３１・・・制
御装置、３２・・・温度検出器、３３・・・圧力検出器
、３５・・・高温ポンプ、３６・・・バイパス配管、３
７・・・流量調節弁、４１・・・温度制御装置、４２・
・・温度検出器、４３・・・圧力検出器、４４・・・圧
力制御装置。FIG. 1 is a block diagram showing one embodiment of the present invention, and FIG.
The figure is a block diagram explaining the conventional technology, Figure 3 is an explanatory diagram of pressure and temperature changes at the time of an accident, Figure 4 is an explanatory diagram of general pump characteristics, flow rate and temperature changes, and Figure 5 is a diagram according to the present invention. An explanatory diagram of control characteristics. 6 is a partial detailed view of FIG. 1, FIG. 7 is a diagram showing the characteristics of FIG. 6, FIG. 8 is a diagram showing the control characteristics of FIG. 1, and FIG. 9 is another embodiment according to the present invention. Block diagram showing the configuration of the first
0 is a block diagram showing the configuration of still another embodiment according to the present invention, FIG. 11 is a diagram explaining the control method of FIG. 10,
FIG. 12 is a diagram showing a basic circuit configuration for achieving temperature control after mixing, and FIG. 13 is a diagram showing a circuit configuration for achieving temperature control in the embodiment shown in FIG. 1...Reactor vessel, 2...Reactor core, 3...Downcomer, 4...Primary coolant, 5...Bot leg, 6...
・Steam generator. 7... Heat exchanger tube, 8... Circulation pump, 9... Cold leg, 10... Pressurizer, 15... Secondary coolant, 2
1... Cooling water storage tank, 22... Cooling water, 23...
Water supply pump, 24... Check valve, 25... Flow control valve, 26... Mixing container. 27... Ejector nozzle, 28... High temperature piping, 29
...Flow control valve, 30...Throat, 31...Control device, 32...Temperature detector, 33...Pressure detector, 35...High temperature pump, 36...Bypass piping, 3
7...Flow control valve, 41...Temperature control device, 42.
...Temperature detector, 43...Pressure detector, 44...Pressure control device.

Claims (1)

【特許請求の範囲】 １、冷却水貯水槽と給水ポンプとを有する原子炉の非常
用炉心冷却方法において、原子炉容器内の一次冷却材を
給水ポンプの下流部に供給し、供給された一次冷却材と
冷却水貯水槽から給水ポンプで供給される冷却水とを混
合するとともに、一次冷却材と冷却水との流量をそれぞ
れ調節して、一次冷却材と冷却水との混合水の温度を冷
却水貯水槽内の圧力に対する飽和温度よりも高くした後
、原子炉容器に供給することを特徴とする原子炉の非常
用炉心冷却方法。 ２、特許請求の範囲第１項において、前記一次冷却材の
流量をまず調節し、一次冷却材の流量調節が限界に達し
た後に、冷却水の流量調節をすることを特徴とする原子
炉の非常用炉心冷却方法。 ３、特許請求の範囲第１項または第２項において、原子
炉容器の圧力が２０気圧以上のときは、混合水の温度を
２００℃以上に調節することを特徴とする原子炉の非常
用炉心冷却方法。 ４、冷却水貯水槽と給水ポンプとを有する原子炉の非常
用炉心冷却装置において、給水ポンプの下流に配置した
混合容器と、一次冷却材容器または一次冷却材配管から
混合容器に一次冷却材を供給する高温配管と、冷却水と
一次冷却材との流量をそれぞれ調節する手段と、混合容
器下流の混合水温度を冷却水貯水槽内の圧力に対する飽
和温度以上にするように前記両流量調節手段を作動させ
る制御装置とを備えたことを特徴とする原子炉の非常用
炉心冷却装置。 ５、特許請求の範囲第４項において、一次冷却水の流量
調節手段が高温配管に設けられた流量調節弁であり、冷
却水の流量調節手段がニードル弁であり、このニードル
弁を含むエゼクタを混合容器内に配置し、エゼクタから
の冷却水の噴出により一次冷却材を吸込み、両者を混合
することを特徴とする原子炉の非常用炉心冷却装置。 ６、特許請求の範囲第４項において、流量調節手段が、
前記冷却水給水ポンプ下流に設けた流量調節弁と、高温
配管に設けた流量調節弁と、高温配管または混合容器下
流に設けた高温ポンプとからなることを特徴とする原子
炉の非常用炉心冷却装置。[Claims] 1. In an emergency core cooling method for a nuclear reactor having a cooling water storage tank and a feed water pump, primary coolant in the reactor vessel is supplied downstream of the feed water pump; The temperature of the mixed water of the primary coolant and cooling water is adjusted by mixing the coolant and the cooling water supplied by the water supply pump from the cooling water storage tank, and adjusting the flow rates of the primary coolant and cooling water respectively. An emergency core cooling method for a nuclear reactor characterized by supplying cooling water to a reactor vessel after raising the temperature to a temperature higher than the saturation temperature for the pressure in a cooling water storage tank. 2. The nuclear reactor according to claim 1, wherein the flow rate of the primary coolant is first adjusted, and after the flow rate adjustment of the primary coolant reaches a limit, the flow rate of the cooling water is adjusted. Emergency core cooling method. 3. An emergency core for a nuclear reactor according to claim 1 or 2, characterized in that when the pressure in the reactor vessel is 20 atm or higher, the temperature of the mixed water is adjusted to 200°C or higher. Cooling method. 4. In an emergency core cooling system for a nuclear reactor that has a cooling water storage tank and a feed water pump, a mixing container placed downstream of the feed water pump and a primary coolant container or primary coolant piping supplying the primary coolant to the mixing container. high-temperature piping to supply, means for adjusting the flow rates of the cooling water and the primary coolant, and means for adjusting both flow rates so that the temperature of the mixed water downstream of the mixing container is equal to or higher than the saturation temperature with respect to the pressure in the cooling water storage tank. An emergency core cooling system for a nuclear reactor, comprising: a control device for operating the system. 5. In claim 4, the primary cooling water flow rate adjustment means is a flow rate adjustment valve provided in the high temperature piping, the cooling water flow rate adjustment means is a needle valve, and an ejector including the needle valve is provided. An emergency core cooling system for a nuclear reactor, which is disposed in a mixing vessel, sucks in primary coolant through a jet of cooling water from an ejector, and mixes both. 6. In claim 4, the flow rate adjusting means comprises:
Emergency core cooling of a nuclear reactor, characterized by comprising a flow rate control valve provided downstream of the cooling water feed pump, a flow rate control valve provided in the high temperature pipe, and a high temperature pump provided downstream of the high temperature pipe or mixing vessel. Device.
JP60069423A 1985-04-02 1985-04-02 Emergency core cooling method and device Expired - Lifetime JPH0713669B2 (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
JP60069423A JPH0713669B2 (en) 1985-04-02 1985-04-02 Emergency core cooling method and device

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
JP60069423A JPH0713669B2 (en) 1985-04-02 1985-04-02 Emergency core cooling method and device

Publications (2)

Publication Number Publication Date
JPS61228391A true JPS61228391A (en) 1986-10-11
JPH0713669B2 JPH0713669B2 (en) 1995-02-15

Family

ID=13402192

Family Applications (1)

Application Number Title Priority Date Filing Date
JP60069423A Expired - Lifetime JPH0713669B2 (en) 1985-04-02 1985-04-02 Emergency core cooling method and device

Country Status (1)

Country Link
JP (1) JPH0713669B2 (en)

Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN103996418A (en) * 2013-02-14 2014-08-20 韩国原子力研究院 Multi stage safety injection device and passive safety injection system having the same

Families Citing this family (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
KR102007252B1 (en) * 2017-12-20 2019-08-06 한국원자력연구원 System for responding to loss of coolant accident

Citations (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPS6021996U (en) * 1983-07-21 1985-02-15 株式会社東芝 Cooling water supply system during reactor isolation

Patent Citations (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPS6021996U (en) * 1983-07-21 1985-02-15 株式会社東芝 Cooling water supply system during reactor isolation

Cited By (2)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN103996418A (en) * 2013-02-14 2014-08-20 韩国原子力研究院 Multi stage safety injection device and passive safety injection system having the same
US9761334B2 (en) 2013-02-14 2017-09-12 Korea Atomic Energy Research Institute Multi stage safety injection device and passive safety injection system having the same

Also Published As

Publication number Publication date
JPH0713669B2 (en) 1995-02-15

Similar Documents

Publication Publication Date Title
JP2012194194A (en) Nuclear reactor
US3981770A (en) Protective arrangements for cooling systems
JPS61228391A (en) Emergency core cooling method and device
US3607636A (en) Nuclear power station with co cooling system
CN104183284B (en) A kind of &#34; passive forced circulation &#34; heat guiding system
JPS5835495A (en) Liquid cooled type reactor
CN109473184A (en) A kind of embedded lead bismuth alloy circuit for fuel irradiation test
JPS61223585A (en) Emergency core cooling device
CN214279618U (en) Nuclear power station cooling system
JPS59143997A (en) Method and system for starting up bwr type reactor
JPH02210295A (en) Auxiliary reactor core cooling device
Dunn Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors
JPS61223584A (en) Emergency pressure-accmulation water-injection facility
CN108413375B (en) Emergency Cooling System and its control method for subcritical circulating fluidized bed boiler
Liu et al. Experimental investigation of early initiation of primary cooldown by secondary-side depressurization in a PWR inadequate core-cooling accident
Chazalon et al. Blanket testing in NET
JPS61167895A (en) Nuclear reactor with control rod with hydraulic driving device
JP3165891B2 (en) Hot air generation amount control method for working fluid heating
JPS6036997A (en) Water supply system for boiling-water type nuclear reactor
JPS62129791A (en) Decay-heat removing operating method
Kukita et al. Simulation of PWR turbine trip transient in ROSA-IV large scale test facility
JPS62200294A (en) Recirculating system of feed pump for nuclear reactor
CN114914001A (en) Nuclear power station cooling system
Paranjpe PFBR reactor protection
JPS63117299A (en) Boiling water type nuclear plant