JPS6383692A - Heat pipe type reactor - Google Patents
Heat pipe type reactorInfo
- Publication number
- JPS6383692A JPS6383692A JP61228155A JP22815586A JPS6383692A JP S6383692 A JPS6383692 A JP S6383692A JP 61228155 A JP61228155 A JP 61228155A JP 22815586 A JP22815586 A JP 22815586A JP S6383692 A JPS6383692 A JP S6383692A
- Authority
- JP
- Japan
- Prior art keywords
- heat pipe
- heat
- nuclear reactor
- airtight container
- reactor according
- 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
- 238000001704 evaporation Methods 0.000 claims description 11
- 230000008020 evaporation Effects 0.000 claims description 11
- QSHDDOUJBYECFT-UHFFFAOYSA-N mercury Chemical compound [Hg] QSHDDOUJBYECFT-UHFFFAOYSA-N 0.000 claims description 8
- 229910052753 mercury Inorganic materials 0.000 claims description 8
- 229910001338 liquidmetal Inorganic materials 0.000 claims description 3
- 238000001816 cooling Methods 0.000 claims 1
- DGAQECJNVWCQMB-PUAWFVPOSA-M Ilexoside XXIX Chemical compound C[C@@H]1CC[C@@]2(CC[C@@]3(C(=CC[C@H]4[C@]3(CC[C@@H]5[C@@]4(CC[C@@H](C5(C)C)OS(=O)(=O)[O-])C)C)[C@@H]2[C@]1(C)O)C)C(=O)O[C@H]6[C@@H]([C@H]([C@@H]([C@H](O6)CO)O)O)O.[Na+] DGAQECJNVWCQMB-PUAWFVPOSA-M 0.000 description 8
- 229910052708 sodium Inorganic materials 0.000 description 8
- 239000011734 sodium Substances 0.000 description 8
- 238000000034 method Methods 0.000 description 3
- 229910000497 Amalgam Inorganic materials 0.000 description 2
- 238000009833 condensation Methods 0.000 description 2
- 230000005494 condensation Effects 0.000 description 2
- 239000000498 cooling water Substances 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 230000000694 effects Effects 0.000 description 2
- 239000007788 liquid Substances 0.000 description 2
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 2
- YREOLPGEVLLKMB-UHFFFAOYSA-N 3-methylpyridin-1-ium-2-amine bromide hydrate Chemical compound O.[Br-].Cc1ccc[nH+]c1N YREOLPGEVLLKMB-UHFFFAOYSA-N 0.000 description 1
- 239000002826 coolant Substances 0.000 description 1
- 230000007797 corrosion Effects 0.000 description 1
- 238000005260 corrosion Methods 0.000 description 1
- 238000001514 detection method Methods 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 238000003466 welding Methods 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
- Devices And Processes Conducted In The Presence Of Fluids And Solid Particles (AREA)
- Physical Or Chemical Processes And Apparatus (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) [Abstract] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
〔産業上の利用分野〕
本発明はヒートパイプを利用して原子炉で発生した熱を
蒸気発生器に伝える装置に関する。DETAILED DESCRIPTION OF THE INVENTION [Field of Industrial Application] The present invention relates to an apparatus for transmitting heat generated in a nuclear reactor to a steam generator using a heat pipe.
高速増殖炉で発生した熱を蒸気発生器に伝える方法とし
て、例えば、特開昭５１−１２９９６３号に示されてい
るようにヒートパイプを用いる方法が考案されている。As a method for transmitting the heat generated in a fast breeder reactor to a steam generator, a method using a heat pipe has been devised, for example, as shown in Japanese Patent Application Laid-Open No. 129963/1983.
第４図の概略図により機能を説明する。第４図において
、炉容器１の炉心２で発生した熱で熱せられた一次ナト
リウム３は炉容器１内にとりつけたヒートパイプ９の蒸
発部４に熱を伝える。この蒸発部４内部では、密封れて
いる熱媒体５が一次ナトリウム３から伝わった熱を受け
、この潜熱をうばうことにより蒸発する。熱媒体５が蒸
発すると、蒸発部４内の圧力が上昇し、熱媒体５の蒸気
はヒートパイプ９により蒸気発生器１０内のヒートパイ
プ９の凝縮部１１へ移動し、ここで蒸気発生器１０内の
冷却水１２に熱をうばわれて凝縮し、液体なる。この凝
縮した熱媒体５は、ヒートパイプ９内を自然落下するか
、あるいはヒートパイプ９を二重管にしたり、多孔質の
ウィック材を内張すしておけば、熱媒体は毛細管現象に
よって蒸発部４に戻る。この方法によると、熱媒体５の
移送にポンプ等の駆動装置を必要とせず、設備が軽減さ
れるとともに、ポンプ等の動力電源が喪失された場合で
も炉心で発生した熱を蒸気発生器に移送でき、かつ炉心
を冷却することができる。ヒートパイプ９は通常何百本
から何千本単位の本数が用いられるともに、熱媒体５に
は通常水銀などの液体金属が用いられる。なお、原子炉
のナトリウム出入口及び蒸気発生器の水／蒸気出入口は
省略しである。１３は原子炉容器カバーガス層、１４は
水蒸気層である。The functions will be explained with reference to the schematic diagram in FIG. In FIG. 4, primary sodium 3 heated by heat generated in the reactor core 2 of the reactor vessel 1 transfers heat to the evaporator section 4 of a heat pipe 9 installed within the reactor vessel 1. Inside this evaporation section 4, a sealed heat medium 5 receives the heat transferred from the primary sodium 3, and evaporates by absorbing this latent heat. When the heat medium 5 evaporates, the pressure inside the evaporation section 4 increases, and the vapor of the heat medium 5 moves through the heat pipe 9 to the condensation section 11 of the heat pipe 9 in the steam generator 10, where it is transferred to the condensation section 11 of the heat pipe 9 in the steam generator 10. The heat is absorbed by the cooling water 12 inside and it condenses, becoming a liquid. This condensed heat medium 5 can fall naturally inside the heat pipe 9, or if the heat pipe 9 is made into a double tube or lined with porous wick material, the heat medium is transferred to the evaporation section by capillary action. Return to 4. According to this method, a driving device such as a pump is not required to transfer the heat medium 5, and the equipment is reduced, and even if the power source for the pump or the like is lost, the heat generated in the reactor core is transferred to the steam generator. and the core can be cooled. The number of heat pipes 9 is usually from hundreds to thousands, and the heat medium 5 is usually a liquid metal such as mercury. Note that the sodium inlet/outlet of the nuclear reactor and the water/steam inlet/outlet of the steam generator are omitted. 13 is a reactor vessel cover gas layer, and 14 is a water vapor layer.
上従来技術は、ヒートパイプ９内の熱媒体５による腐食
などによって生ずるヒートパイプ９の破損に対する安全
設備や破損検出装置に関しては配慮されていなかった。The above conventional technology does not give consideration to safety equipment or a damage detection device for damage to the heat pipe 9 caused by corrosion caused by the heat medium 5 inside the heat pipe 9.
したがって万一ヒートパイプが破損すると熱媒体５が、
原子炉の冷却材中に混入することとなり問題があった。Therefore, in the event that the heat pipe is damaged, the heat medium 5
There was a problem because it got mixed into the coolant of the nuclear reactor.
ヒートパイプの破損は、熱媒体の蒸発部で起こる確率が
高い。There is a high probability that damage to the heat pipe will occur at the evaporation part of the heat medium.
本発明の目的は、万一ヒートパイプが破損しても、熱媒
体５の原子炉内への漏洩を未然に防止する構造を提供す
ることにある。An object of the present invention is to provide a structure that prevents the heat medium 5 from leaking into the nuclear reactor even if the heat pipe is damaged.
ｃ問題点を解決するための手段〕
上記目的は、ヒートパイプ９の蒸発部４を気密容器で保
護する手段を備えて達成される。c. Means for Solving Problems] The above object is achieved by providing means for protecting the evaporation section 4 of the heat pipe 9 with an airtight container.
上記発明によれば、万一ヒートパイプ９が蒸発部４で破
損しても、熱媒体５は気密容器の中に溜り、原子炉内へ
の漏れ込みを防止できる。According to the above invention, even if the heat pipe 9 were to break in the evaporation section 4, the heat medium 5 would remain in the airtight container and would be prevented from leaking into the reactor.
以下、本発明の一実施例を第１図により説明する。ヒー
トパイプ１本をとり出し蒸発部４の近傍を図示したもで
ある。ヒートパイプ９の内側の壁には金属メツシュなど
より成るウィック１５が設けられている。ヒートパイプ
９の下部の蒸発部４には熱媒体５が充填されている。熱
媒体５は通常液体金属を用いられ、一般的には水銀が最
適できる。ヒートパイプ９の下部、すなわち蒸発部４の
媒体５の液面よりややよ方より気密容器１７をかぶせ、
ヒートパイプ９と気密容器１７とを溶接工６で密封する
。気密容器１７にはノズル１８が設けられ、配管工９．
弁２０を介して、それぞれ圧力計２１．ガスクロマトグ
ラフ分析計２２に接続されている。また、気密容器１７
には温度計２３がとりつけられている。熱媒体５として
最適な水銀を高速炉に適用し１本発明の気密容器１７が
ない場合を考えると、水銀は直接炉容器１内のナトリウ
ム中に漏洩することとなり、ナトリウムの水銀アマルガ
ムを作り、以後ナトリウムを精製し水銀アマルガムを除
去しない限りナトリウムの使用は不可能である。このほ
か水銀が蒸気なり、炉容器のカバーガス層１３に溜って
炉容器のガス系により排出され、大気汚染の原因となる
。さて、第１図において、ヒートパイプ９の蒸発部４で
破損が生じた場合、熱媒体５は気密容器７内に溜る。An embodiment of the present invention will be described below with reference to FIG. One heat pipe is taken out and the vicinity of the evaporation section 4 is illustrated. A wick 15 made of metal mesh or the like is provided on the inner wall of the heat pipe 9. The evaporator 4 at the bottom of the heat pipe 9 is filled with a heat medium 5 . The heat medium 5 is usually a liquid metal, and mercury is generally most suitable. An airtight container 17 is placed over the lower part of the heat pipe 9, that is, slightly to the side of the liquid level of the medium 5 in the evaporation section 4.
The heat pipe 9 and the airtight container 17 are sealed by a welder 6. The airtight container 17 is provided with a nozzle 18, and the plumber 9.
Via valves 20, pressure gauges 21. It is connected to a gas chromatograph analyzer 22. In addition, the airtight container 17
A thermometer 23 is attached to the. If mercury, which is optimal as the heat medium 5, is applied to a fast reactor and there is no airtight container 17 of the present invention, mercury will directly leak into the sodium in the reactor container 1, creating a mercury amalgam of sodium. Thereafter, sodium cannot be used unless it is purified and the mercury amalgam removed. In addition, mercury becomes vapor, accumulates in the cover gas layer 13 of the furnace vessel, and is discharged by the gas system of the furnace vessel, causing atmospheric pollution. Now, in FIG. 1, when damage occurs in the evaporation section 4 of the heat pipe 9, the heat medium 5 accumulates in the airtight container 7.
このとき、気密容器内は温度変化を生ずるが、それを温
度計２３で検知することにより、ヒートパイプ９の破損
を知ることができる。また、同時に圧力変化を生ずるの
で、圧力計２１によりそれを検知することができ、さら
に漏洩した熱媒体５のガスクロマトグラフ分析計２２で
分析することにより、同様にヒートパイプ９の破損を検
知することができる。第２図はヒートパイプ９を複数本
、この場合は３本のヒートパイプをブロック化し。At this time, a temperature change occurs inside the airtight container, and by detecting this with the thermometer 23, damage to the heat pipe 9 can be detected. Moreover, since a pressure change occurs at the same time, it can be detected by the pressure gauge 21, and further, by analyzing the leaked heat medium 5 with the gas chromatograph analyzer 22, damage to the heat pipe 9 can be similarly detected. Can be done. Figure 2 shows a block of multiple heat pipes 9, in this case three heat pipes.
３本−緒に気密容器１７にとりつけた例を示す。An example in which three tubes are attached together in an airtight container 17 is shown.
配管１９．弁２０．圧力計２１．ガスクロマトグラフ分
析計２２．温度計２３は省略し図示していないが、基本
的に第１図と同様の構成となる。このようにブロック化
することにより、被損したヒートパイプ１本１本の認定
はできないが、ブロックとして認定することができるの
で実用上の問題はなく、かつ設備としても簡単になり経
済的にも有効である。第３図は第１図及び第２図の発明
に基づき、原子炉へ適用した場合のシステムを図示した
ものである。Piping 19. Valve 20. Pressure gauge 21. Gas chromatograph analyzer 22. Although the thermometer 23 is omitted and not shown, the configuration is basically the same as that in FIG. 1. By creating blocks in this way, it is not possible to recognize each damaged heat pipe individually, but since it can be recognized as a block, there is no practical problem, and the equipment is simple and economical. It is valid. FIG. 3 illustrates a system based on the invention of FIGS. 1 and 2 when applied to a nuclear reactor.
以上本実施例によれば、ヒートバイブロの蒸発部４の破
損によってヒートパイプ９より漏洩する熱媒体５を炉容
器１のナトリウム３内へ混入するのを未然に防止するこ
とができる。そして同時にヒートパイプ９の破損を容易
に検知できるという効果を有する。As described above, according to this embodiment, it is possible to prevent the heat medium 5 leaking from the heat pipe 9 due to breakage of the evaporator section 4 of the heat vibro from mixing into the sodium 3 of the furnace vessel 1. At the same time, damage to the heat pipe 9 can be easily detected.
本発明によれば、ヒートパイプが破損しても安全性を十
分に確保でき、かつ経済性及び効率を大幅に向上できる
という効果がある。According to the present invention, safety can be sufficiently ensured even if the heat pipe is damaged, and economical efficiency and efficiency can be significantly improved.
第１図は本発明の一実施例によるヒートパイプと計測機
器の全体図、第２図は第１図のヒートパイプをブロック
化し簡素化した一実施例を示す断面図、第３図は本発明
を原子炉へ適用した場合のシステムを説明する図、第４
図は従来例の原子炉を示す図である。
１・・・炉容器、２・・・炉心、３・・・ナトリウム、
４・・・蒸発部、５・・・熱媒体、９・・・ヒートパイ
プ、１０・・・蒸気発生器、１１・・・凝縮部、１２・
・・冷却水、１６・・・シール溶接、１７・・・気密容
器、１８・・・ノズル、１９・・・配管、２０・・・弁
、２１・・・圧力計、２２・・・分、−１１、皐１区
光３図Fig. 1 is an overall view of a heat pipe and measuring equipment according to an embodiment of the present invention, Fig. 2 is a sectional view showing an embodiment in which the heat pipe of Fig. 1 is simplified into blocks, and Fig. 3 is a diagram of the present invention. Figure 4 explaining the system when applied to a nuclear reactor.
The figure shows a conventional nuclear reactor. 1... Reactor vessel, 2... Reactor core, 3... Sodium,
4... Evaporation section, 5... Heat medium, 9... Heat pipe, 10... Steam generator, 11... Condensing section, 12.
...Cooling water, 16...Seal welding, 17...Airtight container, 18...Nozzle, 19...Piping, 20...Valve, 21...Pressure gauge, 22...Minute, -11, Go 1-ku Hikari 3 map
Claims (1)
器に熱を伝える原子炉冷却系システムにおいて、前記ヒ
ートパイプを気密容器で保護したことを特徴とするヒー
トパイプ型原子炉。 ２、特許請求の範囲の第１項において、気密容器をヒー
トパイプの蒸発部近傍にとりつけたことを特徴とするヒ
ートパイプ型原子炉。 ３、特許請求の範囲の第１項において、気密容器内に温
度計をとりつけ、かつ気密容器に圧力計及びガス分析装
置を配管と弁と介して接続したことを特徴とするヒート
パイプ型原子炉。 ４、特許請求の範囲の第３項において、ヒートパイプを
複数本ずつブロック化し、そのブロックを気密容器保護
したことを特徴とするヒートパイプ型原子炉。 ５、特許請求範囲の第１項において、ヒートパイプ内に
充填する熱媒体として液体金属を用いたことを特徴とす
るヒートパイプ型原子炉。 ６、特許請求の範囲の第４項において、ヒートパイプ内
に充填する熱媒体として水銀を用いたことを特徴とする
ヒートパイプ型原子炉。[Claims] 1. A nuclear reactor cooling system that transfers heat generated in a nuclear reactor to a heat exchanger using a heat pipe, characterized in that the heat pipe is protected by an airtight container. type reactor. 2. A heat pipe type nuclear reactor according to claim 1, characterized in that an airtight container is attached near the evaporation part of the heat pipe. 3. A heat pipe nuclear reactor according to claim 1, characterized in that a thermometer is installed in an airtight container, and a pressure gauge and a gas analyzer are connected to the airtight container via piping and valves. . 4. A heat pipe nuclear reactor according to claim 3, characterized in that a plurality of heat pipes are formed into blocks and each block is protected in an airtight container. 5. A heat pipe nuclear reactor according to claim 1, characterized in that a liquid metal is used as a heat medium filled in the heat pipe. 6. A heat pipe nuclear reactor according to claim 4, characterized in that mercury is used as a heat medium filled in the heat pipe.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61228155A JPS6383692A (en) | 1986-09-29 | 1986-09-29 | Heat pipe type reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP61228155A JPS6383692A (en) | 1986-09-29 | 1986-09-29 | Heat pipe type reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS6383692A true JPS6383692A (en) | 1988-04-14 |
Family
ID=16872088
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP61228155A Pending JPS6383692A (en) | 1986-09-29 | 1986-09-29 | Heat pipe type reactor |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6383692A (en) |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS64187U (en) * | 1988-06-24 | 1989-01-05 | ||
| JPS643593A (en) * | 1987-06-26 | 1989-01-09 | Power Reactor & Nuclear Fuel | System for utilizing cooling heat of nuclear heated liquid |
| JP2020165836A (en) * | 2019-03-29 | 2020-10-08 | 三菱重工業株式会社 | Nuclear power generation system and nuclear reactor unit |
| CN113345610A (en) * | 2021-05-08 | 2021-09-03 | 中国舰船研究设计中心 | Passive residual heat removal system of heat pipe reactor and control method thereof |
-
1986
- 1986-09-29 JP JP61228155A patent/JPS6383692A/en active Pending
Cited By (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS643593A (en) * | 1987-06-26 | 1989-01-09 | Power Reactor & Nuclear Fuel | System for utilizing cooling heat of nuclear heated liquid |
| JPS64187U (en) * | 1988-06-24 | 1989-01-05 | ||
| JP2020165836A (en) * | 2019-03-29 | 2020-10-08 | 三菱重工業株式会社 | Nuclear power generation system and nuclear reactor unit |
| WO2020202806A1 (en) * | 2019-03-29 | 2020-10-08 | 三菱重工業株式会社 | Nuclear power generation system and nuclear reactor unit |
| CN113345610A (en) * | 2021-05-08 | 2021-09-03 | 中国舰船研究设计中心 | Passive residual heat removal system of heat pipe reactor and control method thereof |
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