JPS59228193A - Nuclear power plant - Google Patents
Nuclear power plantInfo
- Publication number
- JPS59228193A JPS59228193A JP58103928A JP10392883A JPS59228193A JP S59228193 A JPS59228193 A JP S59228193A JP 58103928 A JP58103928 A JP 58103928A JP 10392883 A JP10392883 A JP 10392883A JP S59228193 A JPS59228193 A JP S59228193A
- Authority
- JP
- Japan
- Prior art keywords
- circulation system
- coolant
- sodium
- gas
- steam
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
Landscapes
- Saccharide Compounds (AREA)
- Low-Molecular Organic Synthesis Reactions Using Catalysts (AREA)
- Steroid Compounds (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】
本発明は原子力発電プラントに係シ、特に高速増殖炉を
用いた原子力発電プラントに関するものである。DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a nuclear power plant, and particularly to a nuclear power plant using a fast breeder reactor.
一般に高連中性子を利用した高速増殖炉は、炉内から熱
を取り出す冷却材としてナトリウムを使用している。ナ
トリウムは周知のように沸点が高く、運転温度（１５０
’Ｏ〜６５０ ’Ｏ）で液状を保つので二相流にならな
いという性質を颯
持っており、しかも熱般送能カが大きいため炉内から高
温の熱を取り出すことができ、高速増殖炉の冷却材とし
て適している。In general, fast breeder reactors that utilize high-speed neutrons use sodium as a coolant to extract heat from inside the reactor. As is well known, sodium has a high boiling point and has a high boiling point at operating temperatures (150
It maintains a liquid state at temperatures between 0 and 650 degrees O, so it does not become a two-phase flow, and its large heat transfer capacity allows it to extract high-temperature heat from inside the reactor, making it ideal for fast breeder reactors. Suitable as a coolant.
ところで、このような高速増殖炉を用いて原子力発電を
行う場合、炉内から取り出された高温の冷却材は一般に
中間熱交換器を介して２次冷却材である別のナトリウム
と熱交換され、この２次冷却材と蒸気水とを蒸気発生器
で熱交換させることにより発電用の蒸気タービンを駆動
している。By the way, when nuclear power generation is performed using such a fast breeder reactor, the high temperature coolant taken out from inside the reactor is generally heat exchanged with another sodium, which is a secondary coolant, through an intermediate heat exchanger. A steam turbine for power generation is driven by exchanging heat between this secondary coolant and steam water in a steam generator.
第１図は高速増殖炉を用いた従来の原子力発電プラント
の概略構成図で、図中符号１は高速増殖炉、２は中間熱
交換器、３は蒸気発生器、４は発電機５と直結した蒸気
タービン、６は復水器、７は過水加熱器、８〜１０は循
環ポンプである０即ち、この発電プラントは１次冷却材
であるナトリウムが循環ポンプ８によって高速増殖炉１
と中間熱交換器２との間を循環する１次ナトリウム循環
系１１と、２次冷却材である別のナリウムが循環ポンプ
９によって中間熱交換器２と蒸気発生器３の間を循環す
る２次ナトリウム循環系１２と、蒸気水が循環ボンデ１
゜によって蒸気発生器３と蒸気タービン４の間を復水器
６及び給水加熱器７を介して循環する蒸気水循環系１３
とからなる発電サイクルを形成しており、２次ナトリウ
ム循環系１２を介して蒸気タービン４を駆動している。Figure 1 is a schematic diagram of a conventional nuclear power plant using a fast breeder reactor. In the figure, 1 is a fast breeder reactor, 2 is an intermediate heat exchanger, 3 is a steam generator, and 4 is directly connected to a generator 5. 6 is a condenser, 7 is a superheater, and 8 to 10 are circulation pumps. In other words, in this power plant, sodium, which is the primary coolant, is fed to the fast breeder reactor 1 by the circulation pump 8.
A primary sodium circulation system 11 circulates between the intermediate heat exchanger 2 and the intermediate heat exchanger 2, and another sodium, which is a secondary coolant, is circulated between the intermediate heat exchanger 2 and the steam generator 3 by a circulation pump 9. Sodium circulation system 12 and steam water circulation bonder 1
A steam water circulation system 13 circulates between the steam generator 3 and the steam turbine 4 via the condenser 6 and the feed water heater 7.
The steam turbine 4 is driven via the secondary sodium circulation system 12.
これはナトリウムと水は激しい化学反応を起すので、高
速増殖炉１への影響を考慮して１次す） ＩＪウム循環
系１１と蒸気水循環系１３とを直接熱交換させないため
である。また、この発電プラントは中間熱交換器２が破
損漏えいした場合に放射能汚染された１次冷却拐が２次
ナトリウム循環系１２の中に混入しないように２次冷却
材の圧力を１次冷却拐のそれよシ高く設定している。ま
た、蒸気発生器３での破損漏えいによって２次ナトリウ
ム循環系１２の圧力がナトリウム・水反応によって急上
昇しても大丈夫のように蒸気発生器３の２次ナトリウム
循環系１２側にラゾチャディスク等の圧力開放板１４が
設置されており、圧力の上昇を逃がしている。This is because sodium and water cause a violent chemical reaction, so heat exchange between the IJium circulation system 11 and the steam water circulation system 13 is not allowed. In addition, in this power plant, the pressure of the secondary coolant is reduced to the primary cooling to prevent radioactively contaminated primary cooling waste from entering the secondary sodium circulation system 12 in the event that the intermediate heat exchanger 2 is damaged or leaks. I'm setting it higher than that of the kidnapping. In addition, a lazocha disk etc. are installed on the secondary sodium circulation system 12 side of the steam generator 3 so that it is safe even if the pressure in the secondary sodium circulation system 12 suddenly rises due to a sodium-water reaction due to damage or leakage in the steam generator 3. A pressure release plate 14 is installed to release the pressure increase.
しかしながら、このような原子力発電プラントにおいて
は蒸気発生器３での破損漏えいによるナトリウムと蒸気
水が混り合う危険性を完全に除去することはできない。However, in such a nuclear power plant, it is not possible to completely eliminate the risk that sodium and steam water will mix due to damage and leakage in the steam generator 3.
実際にこ゛のような事態が発生した場合には予想をはる
かに上回る可能性もあり、フ０ラントの健全性が著しく
侵されるおそれがある。If such a situation actually occurs, there is a possibility that it will be much worse than expected, and there is a risk that the soundness of Floating will be seriously damaged.
本発明は上記の事情に鑑みなされたものであり、ナトリ
ウムと水とが混り合う危険性を完全に除去でき、健全性
の高い原子力発電プラントを提供することを目的とする
ものである。The present invention was made in view of the above circumstances, and an object of the present invention is to provide a highly sound nuclear power plant that can completely eliminate the risk of sodium and water mixing.
本発明は上記の目的を達成するために、高速増殖炉の液
体冷却材が循環する１次冷却劇循環系と、上記冷却材と
熱交換関係にある気体冷却拐が循環する気体冷却材循環
系と、上記気体冷却拐によって加熱される蒸気水が循環
する蒸気水循環系とからなる発電サイクルを形成したこ
とを特徴とするものである。In order to achieve the above object, the present invention provides a primary cooling circulation system in which a liquid coolant of a fast breeder reactor circulates, and a gaseous coolant circulation system in which a gaseous cooling liquid having a heat exchange relationship with the coolant circulates. and a steam water circulation system in which steam water heated by the gas cooling is circulated.
以下、図面を参照して本発明の実施例を詳細に説明する
。Embodiments of the present invention will be described in detail below with reference to the drawings.
第２図は本発明の一実施例を示す原子力発電プラントの
概略構成図で、図中第１図と同一部分には同一符号が付
されている。同図シζ示すようにこの原子発電プラント
は、高速増殖炉１の液体冷却材であるナトリウムが循環
ボンデ８によって高速増殖炉１と中間熱交換器２との間
を循環する１次ナトリウム循環系１ノと、２次冷却材と
しての気体冷力１材が循環ファン２１によって中間熱交
換器２と号？イラ２２との間を循環する気体冷却材循環
系２ｏと、蒸気水が循環ボンデ１０によってｓｐイラ２
２と蒸気タービン４との間を復水器６及び給水加熱器７
を介し７て循環する蒸気水循環系１３とからなる発電サ
イクルを形成しており、気体冷却材循環系２ｏを介して
蒸気タービン４を駆動している。この気体冷却材循環系
２ｏの気体冷却↓」とし７て（６シ例えばヘリウムガス
、窒素力′ス、炭酸ガス等が使用可能であり、高速増殖
炉１のシールガス（例えばアルゴンガス）より＠い気体
である。また、気体冷却材循環系２ｏの圧力は１１次す
）ＩＪウム循環系１１の圧力よシ若干高く設定されてお
り、しかも蒸気水循環系１３のそｔｌよ（ン低く設定さ
れている。FIG. 2 is a schematic configuration diagram of a nuclear power plant showing an embodiment of the present invention, and the same parts as in FIG. 1 are given the same reference numerals. As shown in figure ζ, this nuclear power plant has a primary sodium circulation system in which sodium, which is the liquid coolant of the fast breeder reactor 1, is circulated between the fast breeder reactor 1 and the intermediate heat exchanger 2 by a circulation bonder 8. 1 and gas cooling power 1 as a secondary coolant are transferred to an intermediate heat exchanger 2 by a circulation fan 21. A gas coolant circulation system 2o circulates between the spiller 22 and the steam water is passed through the circulation bonder 10 to the spiraer 2.
A condenser 6 and a feed water heater 7 are connected between the steam turbine 2 and the steam turbine 4.
A power generation cycle is formed with a steam water circulation system 13 that circulates through a gas coolant circulation system 2o, and a steam turbine 4 is driven via a gas coolant circulation system 2o. For example, helium gas, nitrogen gas, carbon dioxide gas, etc. can be used for gas cooling of this gas coolant circulation system 2o, and the seal gas (for example, argon gas) of the fast breeder reactor 1 can be used. In addition, the pressure of the gas coolant circulation system 2o is set slightly higher than that of the 11th-order coolant circulation system 11, and is set lower than that of the steam water circulation system 13. ing.
従って、高速増殖炉１から取（）出された高温のナトリ
ウノ・句却拐＆」中間省！２文換器２で気体冷却材循環
系２０の気体冷却制とＰ、交換され、低温となって再び
高速贈殖炉ノに循＋Ｊ１（ポンプ１θによって送り込ま
れる。一方、中間熱交換器２で高温のナトリウム冷却利
と熱交換器た気体冷却材は高温となり、ボイラ２２で蒸
気水循環系１３の蒸気水と熱交換さハ１、低温となって
再び循環ファン２１によって中旬黙蝉博２に送り込まれ
る。また、ボイラ２２で高温の気体冷却祠と熱交換した
蒸気水は加熱さハ過熱蒸気となり、発電枳５と結合した
蒸気タービン４に送られてタービン駆動する。そして、
蒸気タービン４で熱の質された蒸気水は循環ポンプ１０
によって再び復水器６及び給水加熱器２を経てボイラ２
２に送り込まね、過熱蒸気となる。Therefore, the high-temperature sodium chloride taken out from fast breeder reactor 1 was rejected &'' Intermediate Ministry! In the exchanger 2, P is exchanged with the gas cooling system of the gas coolant circulation system 20, and the low temperature is circulated again to the high speed feeding reactor (Pump 1θ).Meanwhile, in the intermediate heat exchanger 2 The high-temperature sodium coolant and the gaseous coolant in the heat exchanger become high temperature, and are exchanged with steam water in the steam water circulation system 13 in the boiler 22.Then, the temperature becomes low and is sent to the mid-mokusemi-hiro 2 by the circulation fan 21 again. In addition, the steam water that has exchanged heat with the high-temperature gas cooling plant in the boiler 22 is heated and becomes superheated steam, which is sent to the steam turbine 4 connected to the power generating shaft 5 to drive the turbine.
The steam water heated by the steam turbine 4 is sent to the circulation pump 10.
The boiler 2 passes through the condenser 6 and the feed water heater 2 again.
2, it becomes superheated steam.
このように本実施例においては、高速垢殖炉Ｊの１次ナ
トリウム循環系１１と熱交換関係にある２次冷却材循環
系を気体冷却材循環系２０としたので、ナトリウムと水
とが混り合う危険性を完全に除去することが可能である
。また、気体冷却材循環系２０の気体冷却材は高速増殖
炉１のシールガスよりも軽い気体であるので、たとえ中
間熱交換器２の破損漏えいによって１次ナトリウム循環
系１１の中に混入しても高速増殖炉１のシールガスより
上に抜け、しかも１次ナトリウム循環系１１よりも圧力
ガスが若干高く設定されているのでナトリウム冷却材が
気体冷却材循環系２０の中に混入することはない０なお
、上記実施例では蒸気水循環系１３として単玉サイクル
のものを示したが、第３図及び第４図ｊＣ示すように単
玉再熱サイクルや複圧再熱サイクルのものでも実施可能
であり、このようにすれば熱効率の高い発電サイクルを
形成することができる。In this way, in this embodiment, the gas coolant circulation system 20 is used as the secondary coolant circulation system which is in a heat exchange relationship with the primary sodium circulation system 11 of the fast fermentation furnace J, so that sodium and water are mixed. It is possible to completely eliminate the risk of collision. In addition, since the gas coolant in the gas coolant circulation system 20 is a lighter gas than the seal gas of the fast breeder reactor 1, even if the intermediate heat exchanger 2 is damaged and leaks, it may get mixed into the primary sodium circulation system 11. The sodium coolant also escapes above the seal gas of the fast breeder reactor 1, and the pressure gas is set slightly higher than that of the primary sodium circulation system 11, so sodium coolant does not mix into the gas coolant circulation system 20. 0 In the above embodiments, a single bulb cycle was shown as the steam water circulation system 13, but it can also be implemented with a single bulb reheat cycle or a double pressure reheat cycle, as shown in Figures 3 and 4 jC. In this way, a power generation cycle with high thermal efficiency can be formed.
以上述べたように本発明によれば、高速増殖炉の液体冷
却材が循環する１次冷却材循環系と上記冷却材と熱交換
関係にある気体冷却材が循健全性の高い原子力発電プラ
ントを提供できる０As described above, according to the present invention, the primary coolant circulation system in which the liquid coolant of the fast breeder reactor circulates and the gas coolant that is in a heat exchange relationship with the coolant generate a nuclear power plant with high circulation integrity. 0 that can be provided
第１図は従来の原子力発電プラントの概略構成図、第２
図は本発明の一実施例を示す原子力発電プラントの概略
構成図、第３図は単玉再熱サイクルの蒸気水循環系の場
合を示す概略構成図、第４図は複圧再熱サイクルの蒸気
水循環系の場合を示す概略構成図である。
１・・・高速増殖炉、２・・・中間熱交換器、４・・・
蒸気タービン、５・・・発電機、６・・・復水器、７・
・・給水加熱器、８，１０・・・循環ポンプ、１ノ・・
・１次ナトリウム循環系、１３・・・蒸気水循環系、２
０・・・気体冷却材循環系、２ノ・・・循環ファン、２
２・・・ボイラ。Figure 1 is a schematic diagram of a conventional nuclear power plant;
The figure is a schematic block diagram of a nuclear power plant showing an embodiment of the present invention, Figure 3 is a schematic block diagram showing the case of a steam water circulation system in a single-bulk reheat cycle, and Figure 4 is a schematic block diagram showing a steam water circulation system in a double pressure reheat cycle. It is a schematic block diagram showing the case of a water circulation system. 1... Fast breeder reactor, 2... Intermediate heat exchanger, 4...
Steam turbine, 5... Generator, 6... Condenser, 7.
...Water heater, 8,10...Circulation pump, 1...
・Primary sodium circulation system, 13...Steam water circulation system, 2
0...Gas coolant circulation system, 2No...Circulation fan, 2
2...Boiler.
Claims (1)
よって加熱される蒸気水が循環する蒸気水循環系とから
なる発電サイクルを形成したことを特徴とする原子力発
電プラント。A nuclear power generation plant characterized by forming a power generation cycle consisting of a steam water circulation system in which steam water heated by a primary coolant coolant in which a liquid coolant of a fast breeder reactor circulates.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58103928A JPS59228193A (en) | 1983-06-10 | 1983-06-10 | Nuclear power plant |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP58103928A JPS59228193A (en) | 1983-06-10 | 1983-06-10 | Nuclear power plant |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS59228193A true JPS59228193A (en) | 1984-12-21 |
Family
ID=14367088
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP58103928A Pending JPS59228193A (en) | 1983-06-10 | 1983-06-10 | Nuclear power plant |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS59228193A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS61178505A (en) * | 1985-02-02 | 1986-08-11 | クラウス クニチア | Method and device for operating power plant |
-
1983
- 1983-06-10 JP JP58103928A patent/JPS59228193A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS61178505A (en) * | 1985-02-02 | 1986-08-11 | クラウス クニチア | Method and device for operating power plant |
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