JPS6111395B2 - - Google Patents
Info
- Publication number
- JPS6111395B2 JPS6111395B2 JP55008643A JP864380A JPS6111395B2 JP S6111395 B2 JPS6111395 B2 JP S6111395B2 JP 55008643 A JP55008643 A JP 55008643A JP 864380 A JP864380 A JP 864380A JP S6111395 B2 JPS6111395 B2 JP S6111395B2
- Authority
- JP
- Japan
- Prior art keywords
- reactor
- reactor vessel
- water
- main steam
- pressure vessel
- 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.)
- Expired
Links
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 24
- 238000001816 cooling Methods 0.000 claims description 17
- 239000002826 coolant Substances 0.000 claims description 11
- 239000000498 cooling water Substances 0.000 description 20
- 238000002955 isolation Methods 0.000 description 10
- 238000010586 diagram Methods 0.000 description 6
- 238000009835 boiling Methods 0.000 description 4
- 238000000746 purification Methods 0.000 description 4
- 238000007689 inspection Methods 0.000 description 3
- 238000012423 maintenance Methods 0.000 description 3
- 239000007921 spray Substances 0.000 description 3
- 239000000428 dust Substances 0.000 description 2
- 230000005855 radiation Effects 0.000 description 2
- 230000001172 regenerating effect Effects 0.000 description 2
- 230000007797 corrosion Effects 0.000 description 1
- 238000005260 corrosion Methods 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 239000000446 fuel Substances 0.000 description 1
- 239000007791 liquid phase Substances 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 230000035515 penetration Effects 0.000 description 1
- 239000012071 phase Substances 0.000 description 1
- 230000010363 phase shift Effects 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Landscapes
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Description
【発明の詳細な説明】
本発明は原子炉容器の冷却設備に係り、特に沸
騰水型原子炉停止時に原子炉圧力容器を冷却する
のに好適な原子炉の冷却設備に関するものであ
る。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to cooling equipment for a nuclear reactor vessel, and particularly to a cooling equipment for a nuclear reactor suitable for cooling a reactor pressure vessel during shutdown of a boiling water reactor.
沸騰水型原子炉を停止させ、原子炉の燃料交換
等の保守点検のため、原子炉圧力容器の上蓋を取
り外す場合は、炉水を冷却しながら原子炉圧力容
器自体を冷却する必要がある。そのため、従来は
第1図に示すように、原子炉圧力容器1の上蓋2
を取り外す前に、原子炉圧力容器1内の炉水(一
次冷却水)3を強制循環する再循環ポンプ4を有
する再循環系5に並設したポンプ6、熱交換器7
などよりなる残留熱除去系8にポンプ6により一
次冷却水3を取り出し、熱交換器7で冷却し、そ
れを上蓋2のスプレイノズル9より原子炉圧力容
器1内に封入するようにし、再循環系5からの一
次冷却水3の注入とともに、炉水および原子炉圧
力容器1自体の冷却を行うようにしていた。とこ
ろで、原子炉圧力容器1自体の冷却を行う場合、
原子炉圧力容器1内の一次冷却水3の水位を上蓋
2のところまで高めるようにすると、原子炉圧力
容器1を一次冷却水3と同一温度まで効率的に冷
却できる。しかし、第1図の場合は、そのように
すると、原子炉圧力容器1で発生する蒸気を蒸気
タービン(図示せず)に導く主蒸気配管10に一
次冷却水3が流入し、これにともない、一次冷却
水3の水面に浮遊する金属腐食生成物(クラツ
ド)11が侵入し、逃し安全弁12および主蒸気
隔離弁13Aおよび13Bにいわゆるごみが発生
し、また、主蒸気配管10の放射線量率の上昇を
招く原因になる。そのため、一次冷却水3の水位
を主蒸気配管10の位置ぎりぎりとして上記によ
る冷却をするようにしていた。 When shutting down a boiling water reactor and removing the top cover of the reactor pressure vessel for maintenance inspections such as reactor fuel replacement, it is necessary to cool the reactor pressure vessel itself while cooling the reactor water. Therefore, conventionally, as shown in FIG. 1, the upper cover 2 of the reactor pressure vessel 1
Before removing the reactor pressure vessel 1, a pump 6 and a heat exchanger 7 are installed in parallel to a recirculation system 5 having a recirculation pump 4 that forcibly circulates reactor water (primary cooling water) 3 in the reactor pressure vessel 1.
The primary cooling water 3 is taken out by a pump 6 into a residual heat removal system 8 consisting of a residual heat removal system 8, cooled by a heat exchanger 7, and then sealed into the reactor pressure vessel 1 through a spray nozzle 9 of the upper cover 2, and then recirculated. In addition to injecting the primary cooling water 3 from the system 5, the reactor water and the reactor pressure vessel 1 itself were cooled. By the way, when cooling the reactor pressure vessel 1 itself,
By raising the water level of the primary cooling water 3 in the reactor pressure vessel 1 to the level of the upper cover 2, the reactor pressure vessel 1 can be efficiently cooled to the same temperature as the primary cooling water 3. However, in the case of FIG. 1, if this is done, the primary cooling water 3 will flow into the main steam pipe 10 that guides the steam generated in the reactor pressure vessel 1 to the steam turbine (not shown), and as a result, Metal corrosion products (cruds) 11 floating on the water surface of the primary cooling water 3 enter, so-called dust is generated in the safety relief valve 12 and the main steam isolation valves 13A and 13B, and the radiation dose rate of the main steam piping 10 is reduced. It will cause an increase. For this reason, the water level of the primary cooling water 3 is set to the level of the main steam pipe 10 to perform the above-mentioned cooling.
したがつて、原子炉圧力容器1の一次冷却水3
に接触していない部分の冷却は、内部雰囲気の自
然循環で行われることになり、原子炉圧力容器1
全体を冷却するのに長時間を要し、他の作業に移
る時間的制約条件になつていた。また、原子炉圧
力容器1の冷却時に原子炉圧力容器1の液相部分
と気相部分とで大きな温度差を生じ、それにより
発生する熱ひずみが問題になつていた。 Therefore, the primary cooling water 3 of the reactor pressure vessel 1
Cooling of the parts not in contact with the reactor pressure vessel 1 will be performed by natural circulation of the internal atmosphere.
It took a long time to cool down the whole thing, which became a time constraint for moving on to other tasks. Further, when the reactor pressure vessel 1 is cooled, a large temperature difference occurs between the liquid phase portion and the gas phase portion of the reactor pressure vessel 1, and the resulting thermal strain has become a problem.
本発明は上記に鑑みてなされたもので、その目
的とするところは、原子炉停止時に冷却材中のク
ラツドを主蒸気配管内に侵入させることなく原子
炉容器を効率的に冷却することができる原子炉容
器の冷却設備を提供することにある。 The present invention has been made in view of the above, and its purpose is to efficiently cool the reactor vessel without allowing crud in the coolant to enter the main steam piping when the reactor is shut down. Its purpose is to provide cooling equipment for nuclear reactor vessels.
本発明の第1の特徴は、原子炉容器内で発生す
る蒸気を蒸気タービンに導く主蒸気管より冷却材
を上記原子炉容器に供給する手段を具備させた点
にある。第2の特徴は、主蒸気管より原子炉一次
冷却水の強制循環径路に並設した残留熱除去系の
冷却材を取り出して浄化、冷却した冷却材を原子
炉容器に供給する手段と、他の系統の冷却材を上
記原子炉容器に供給して充水する手段とを具備さ
せた点にある。 A first feature of the present invention resides in that a means is provided for supplying coolant to the reactor vessel from a main steam pipe that guides steam generated within the reactor vessel to a steam turbine. The second feature is a means for extracting the coolant from the residual heat removal system installed in parallel with the forced circulation path of the reactor primary cooling water from the main steam pipe, purifying it, supplying the cooled coolant to the reactor vessel, and other features. The reactor vessel is provided with means for supplying a coolant of the system to the reactor vessel to fill it with water.
以下、沸騰水型原子炉に適用した本発明を第2
図、第3図に示した実施例を用いて詳細に説明す
る。 Hereinafter, the present invention applied to a boiling water reactor will be explained in the second section.
This will be explained in detail using the embodiment shown in FIGS.
第2図は本発明の冷却設備の一実施例を示す系
統図である。ただし、第1図と同一部分は同じ符
号で示し、説明を省略する。第2図においては、
残留熱除去系8より分岐して設けた配管15、循
環ポンプ16、非再生熱交換器17、一次冷却水
浄化装置18、再生熱交換器19よりなる原子炉
浄化系20が原子炉給水配管21に合流する部分
の弁22の手前で分岐された連絡配管23を設
け、連絡配管23を格納容器14の外側で主蒸気
隔離弁13Aと主蒸気隔離弁13Bとの間の主蒸
気配管10に接続する。連絡配管23を格納容器
14の外側で主蒸気配管10に接続しているの
で、格納容器14を貫通して連絡配管23を配置
する必要がなく、格納容器14に配管貫通部を新
たに設ける必要がない。 FIG. 2 is a system diagram showing an embodiment of the cooling equipment of the present invention. However, the same parts as in FIG. 1 are indicated by the same reference numerals, and their explanation will be omitted. In Figure 2,
A reactor purification system 20 consisting of a piping 15 branched from the residual heat removal system 8, a circulation pump 16, a non-regenerative heat exchanger 17, a primary cooling water purification device 18, and a regenerative heat exchanger 19 is connected to the reactor water supply piping 21. A connecting pipe 23 is provided which is branched before the valve 22 at the part where it joins the main steam pipe 10, and the connecting pipe 23 is connected to the main steam pipe 10 between the main steam isolation valve 13A and the main steam isolation valve 13B outside the containment vessel 14. do. Since the communication pipe 23 is connected to the main steam pipe 10 outside the containment vessel 14, there is no need to arrange the communication pipe 23 through the containment vessel 14, and it is not necessary to newly provide a pipe penetration part in the containment vessel 14. There is no.
そして、原子炉停止時に、燃料交換等の保守点
検のため、原子炉圧力容器1および炉水3を冷却
するときは、再循環系5を用いて一次冷却水3を
強制循環により注入するとともに、残留熱除去系
8を通して上蓋2のスプレイノズル9より一次冷
却水3を注入し、さらに、原子炉浄化系20から
連絡配管23、主蒸気配管10を通して、原子炉
浄化系20で浄化、冷却された一次冷却水3を注
入する。この時、主蒸気隔離弁13Bは、冷却水
がタービンに供給されることを防止するために、
閉鎖されている。主蒸気隔離弁13Aは、開いて
いる。また、この状態で原子炉給水配管21より
復水貯蔵タンク(図示せず)内の脱塩水(一次冷
却水)を原子炉圧力容器1内に供給し、原子炉圧
力容器1内を満水にするようにする。 When the reactor is shut down and the reactor pressure vessel 1 and reactor water 3 are to be cooled for maintenance inspections such as fuel exchange, the primary cooling water 3 is injected by forced circulation using the recirculation system 5, and Primary cooling water 3 is injected from the spray nozzle 9 of the upper lid 2 through the residual heat removal system 8, and is further purified and cooled by the reactor purification system 20 through the connecting pipe 23 and the main steam pipe 10. Inject primary cooling water 3. At this time, the main steam isolation valve 13B prevents cooling water from being supplied to the turbine.
Closed. Main steam isolation valve 13A is open. In addition, in this state, desalinated water (primary cooling water) in the condensate storage tank (not shown) is supplied from the reactor water supply pipe 21 to the reactor pressure vessel 1 to fill the reactor pressure vessel 1 with water. Do it like this.
このようにすれば、原子炉圧力容器1内を満水
にするときに、炉水3の面に浮遊するクラツド1
1が主蒸気配管10に流れ込むことがなく、逃し
安全弁12および主蒸気隔離弁13Aおよび13
Bにいわゆるごみかみが発生したり、主蒸気配管
10の放射線量率の上昇を招くことがない。ま
た、原子炉圧力容器1内を一次冷却水3で満水に
するので、原子炉圧力容器1が均一な温度に冷却
され、原子炉圧力容器1に熱ひずみが発生する問
題がなくなる。しかも、冷却時間を著しく短縮す
ることができる。 In this way, when the inside of the reactor pressure vessel 1 is filled with water, the crud 1 floating on the surface of the reactor water 3 can be removed.
1 does not flow into the main steam piping 10, and the relief safety valve 12 and the main steam isolation valves 13A and 13
There is no possibility that so-called dust traps will occur in B, and the radiation dose rate of the main steam pipe 10 will not increase. Furthermore, since the reactor pressure vessel 1 is filled with primary cooling water 3, the reactor pressure vessel 1 is cooled to a uniform temperature, and the problem of thermal strain occurring in the reactor pressure vessel 1 is eliminated. Moreover, the cooling time can be significantly shortened.
このため、原子炉圧力容器1の上蓋2を早くあ
けることができ、保守点検に早く着手することが
できる。これは、沸騰水型原子炉の停止期間を短
縮させ、稼動率を向上させることにつながる。 Therefore, the upper cover 2 of the reactor pressure vessel 1 can be opened quickly, and maintenance inspection can be started quickly. This will shorten the shutdown period of boiling water reactors and improve their operating rates.
第3図は本発明の他の実施例を示す系統図で、
第1図と同一部分は同じ符号で示してある。第3
図においては、主蒸気配管10に設けた主蒸気隔
離弁13Aと主蒸気隔離弁13Bとの間の主蒸気
配管10に復水貯蔵タンク24内の脱塩水を移送
ポンプ25で加熱器26を通して導びくようにし
それを主蒸気隔離弁13A、主蒸気配管10を経
て原子炉圧力容器1に供給し、原子炉圧力容器1
を満水にするようにした。この時、主蒸気隔離弁
13Bは、閉じている。なお、この場合、図示を
省略してあるが、再循環系5のほか第1図に示し
た残留熱除去系8から一次冷却水3を注入するこ
とは同様であり、なお、加熱器26は復水貯蔵タ
ンク24からの脱塩水の温度を一次冷却水3と同
程度とするために設けた。第3図のようにして
も、その効果は第2図の場合と同様である。 FIG. 3 is a system diagram showing another embodiment of the present invention,
The same parts as in FIG. 1 are designated by the same reference numerals. Third
In the figure, desalinated water in a condensate storage tank 24 is introduced through a heater 26 by a transfer pump 25 to the main steam pipe 10 between the main steam isolation valve 13A and the main steam isolation valve 13B provided in the main steam pipe 10. It is supplied to the reactor pressure vessel 1 via the main steam isolation valve 13A and the main steam piping 10, and the reactor pressure vessel 1
I made sure to fill it with water. At this time, the main steam isolation valve 13B is closed. In this case, although not shown, the primary cooling water 3 is injected from the residual heat removal system 8 shown in FIG. 1 in addition to the recirculation system 5, and the heater 26 is This was provided in order to make the temperature of the desalinated water from the condensate storage tank 24 comparable to that of the primary cooling water 3. Even if the configuration is as shown in FIG. 3, the effect is the same as in the case of FIG.
以上説明したように、本発明によれば、原子炉
停止時に冷却材中のクラツドを主蒸気配管内に侵
入させることなく、原子炉容器を効率的に冷却で
きるという効果がある。 As explained above, according to the present invention, the reactor vessel can be efficiently cooled without allowing crud in the coolant to enter the main steam piping when the reactor is shut down.
第1図は従来の原子炉圧力容器の冷却設備の系
統図、第2図は本発明の原子炉圧力容器の冷却設
備の一実施例を示す系統図、第3図は本発明の他
の実施例を示す系統図である。
1……原子炉圧力容器、2……上蓋、3……一
次冷却水、5……再循環系、8……残留熱除去
系、9……スプレイノズル、10……主蒸気配
管、20……原子炉浄化系、21……原子炉給水
配管、23……連絡配管、24……復水貯蔵タン
ク、25……移相ポンプ、26……加熱器。
Fig. 1 is a system diagram of a conventional reactor pressure vessel cooling equipment, Fig. 2 is a system diagram showing an embodiment of the reactor pressure vessel cooling equipment of the present invention, and Fig. 3 is a system diagram showing another embodiment of the reactor pressure vessel cooling equipment of the present invention. It is a system diagram showing an example. DESCRIPTION OF SYMBOLS 1...Reactor pressure vessel, 2...Upper lid, 3...Primary cooling water, 5...Recirculation system, 8...Residual heat removal system, 9...Spray nozzle, 10...Main steam piping, 20... ... Reactor purification system, 21 ... Reactor water supply pipe, 23 ... Connection pipe, 24 ... Condensate storage tank, 25 ... Phase shift pump, 26 ... Heater.
Claims (1)
おいて、該原子炉容器内で発生する蒸気を蒸気タ
ービンに導く主蒸気管より冷却材を前記原子炉容
器に供給する手段を具備していることを特徴とす
る原子炉容器の冷却設備。 2 前記手段が復水貯蔵タンク内の冷却材を原子
炉容器に供給して充水するものである特許請求の
範囲第1項記載の原子炉容器の冷却設備。 3 原子炉停止時に原子炉容器を冷却するものに
おいて、該原子炉容器内で発生する蒸気を蒸気タ
ービンに導く主蒸気管より原子炉一次冷却材の強
制循環径路に並設した残留熱除去系の冷却材を取
り出して浄化、冷却した冷却材を前記原子炉容器
に供給する手段と、他の系統の冷却材を前記原子
炉容器に供給して充水する手段とを具備している
ことを特徴とする原子炉容器の冷却設備。[Scope of Claims] 1. A device for cooling a reactor vessel when a nuclear reactor is shut down, comprising means for supplying coolant to the reactor vessel from a main steam pipe that guides steam generated within the reactor vessel to a steam turbine. Cooling equipment for a nuclear reactor vessel characterized by comprising: 2. The reactor vessel cooling equipment according to claim 1, wherein the means supplies the coolant in the condensate storage tank to the reactor vessel to fill it with water. 3. In systems that cool the reactor vessel during reactor shutdown, a residual heat removal system installed in parallel with the forced circulation path of the reactor primary coolant from the main steam pipe that guides the steam generated in the reactor vessel to the steam turbine. It is characterized by comprising means for supplying the reactor vessel with coolant that has been extracted, purified and cooled, and means for supplying coolant from another system to the reactor vessel to fill it with water. cooling equipment for the reactor vessel.
Priority Applications (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP864380A JPS56106189A (en) | 1980-01-28 | 1980-01-28 | Cooling facility of nuclear reactor container |
Applications Claiming Priority (1)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
JP864380A JPS56106189A (en) | 1980-01-28 | 1980-01-28 | Cooling facility of nuclear reactor container |
Publications (2)
Publication Number | Publication Date |
---|---|
JPS56106189A JPS56106189A (en) | 1981-08-24 |
JPS6111395B2 true JPS6111395B2 (en) | 1986-04-02 |
Family
ID=11698624
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
JP864380A Granted JPS56106189A (en) | 1980-01-28 | 1980-01-28 | Cooling facility of nuclear reactor container |
Country Status (1)
Country | Link |
---|---|
JP (1) | JPS56106189A (en) |
Cited By (1)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPH0345796U (en) * | 1989-09-11 | 1991-04-26 |
-
1980
- 1980-01-28 JP JP864380A patent/JPS56106189A/en active Granted
Cited By (1)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPH0345796U (en) * | 1989-09-11 | 1991-04-26 |
Also Published As
Publication number | Publication date |
---|---|
JPS56106189A (en) | 1981-08-24 |
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