JPS61228391A - Emergency core cooling method and device - Google Patents
Emergency core cooling method and deviceInfo
- 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
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.
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).
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.
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.
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.
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)
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.
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)
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)
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)
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 |
-
1985
- 1985-04-02 JP JP60069423A patent/JPH0713669B2/en not_active Expired - Lifetime
Patent Citations (1)
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)
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 |
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